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Sample records for graphite irradiation creep

  1. Irradiation creep of graphite

    International Nuclear Information System (INIS)

    Kennedy, C.R.

    1990-01-01

    Displacement damage of graphite by neutron irradiation causes graphite to change dimensions. This dimensional instability requires careful attention when graphite is used as as moderator and reflector material in nuclear devices. Natural gradients in flux and temperature result in time-varying differential growth generating stresses similar to thermal stresses with an ever increasing temperature gradient. Graphite, however, does have the ability to creep under irradiation, allowing the stress intensity to relax below the fracture strength of the material. Creep strain also serves to average the radiation-induced strains, thus contributing to the stability of the core. As the dimensional instability is a function of temperature, so are the creep characteristics of graphite, and it is of interest to generalize the available data for extension to more extreme conditions of fluence and temperature. Irradiation creep of graphite is characterized by two stages of creep; a primary stage that saturates with time and a secondary stage that is generally assumed to be linear and constant with time. Virtually all past studies have not considered primary creep in detail primarily because there is limited available data at the very low fluences required to saturate primary creep. It is the purpose of this study to carefully examine primary creep in detail over the irradiation temperature range of 150 to 1000 degree C. These studies also include the combined effects of creep, differential growth, and structural changes in graphite by irradiation. 3 refs., 5 figs

  2. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  3. (Irradiation creep of graphite)

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, C.R.

    1990-12-21

    The traveler attended the Conference, International Symposium on Carbon, to present an invited paper, Irradiation Creep of Graphite,'' and chair one of the technical sessions. There were many papers of particular interest to ORNL and HTGR technology presented by the Japanese since they do not have a particular technology embargo and are quite open in describing their work and results. In particular, a paper describing the failure of Minor's law to predict the fatigue life of graphite was presented. Although the conference had an international flavor, it was dominated by the Japanese. This was primarily a result of geography; however, the work presented by the Japanese illustrated an internal program that is very comprehensive. This conference, a result of this program, was better than all other carbon conferences attended by the traveler. This conference emphasizes the need for US participation in international conferences in order to stay abreast of the rapidly expanding HTGR and graphite technology throughout the world. The United States is no longer a leader in some emerging technologies. The traveler was surprised by the Japanese position in their HTGR development. Their reactor is licensed and the major problem in their graphite program is how to eliminate it with the least perturbation now that most of the work has been done.

  4. Significance of primary irradiation creep in graphite

    CSIR Research Space (South Africa)

    Erasmus, C

    2013-05-01

    Full Text Available Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux...

  5. Irradiation creep performance of graphite relevant for pebble bed HTRs

    International Nuclear Information System (INIS)

    Kleist, G.; O'Connor, M.F.

    1980-01-01

    Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

  6. The irradiation creep characteristics of graphite to high fluences

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Cundy, M.; Kleist, G.

    1988-01-01

    High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

  7. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  8. A review of irradiation induced creep in graphite under CAGR conditions

    International Nuclear Information System (INIS)

    Brocklehurst, J.E.; Kelly, B.T.

    1989-06-01

    Graphite irradiation induced creep data have been reviewed in detail and compared with the existing model used for stress calculations under Commercial Advanced Gas Cooled Reactor conditions. The relationship between creep and elastic modulus is well established and allows the creep behaviour of any graphite in any orientation to be predicted. The model predicts the initial build up of creep strain in different graphites extremely well. However, there are differences between prediction and experiment in creep at high doses; a creep test on a pre-irradiated specimen showed rather more creep ductility than predicted, whilst in experiments under a constant stress applied for the whole period of irradiation, the creep rates decreased to lower values than predicted. It is suggested that to obtain better estimates of brick stresses future irradiation experiments should aim to generate creep data under realistic variations in applied stress level, or rather creep strain history. (author)

  9. A Review of Graphite Irradiation Creep Data from the "OC-Series" of Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Davies, Mark A. [MARAD Co. Ltd., Washington, DC (United States); Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2012-09-01

    The OC-Series graphite irradiation creep experiments were conducted in the early 1970s in the Oak Ridge Research Reactor (ORR) at ORNL. The OC Series consisted of 5 experiments, Capsules 1, 3 and 5 were irradiated at 900°C and Capsules 2 and 4 were irradiated at 600°C. Each capsule contained four columns of specimens, two loaded in compression and two un-loaded. The loaded columns had specimens of different diameter to generate two stress levels, 13.8 MPa and 20.7 MPa. Some of the data from these experiments were presented in extended abstracts at a Carbon Conference (Kennedy et al, 1977: Kennedy and Eatherly, 1979). The data presented some challenges to the accepted approach to irradiation induced creep in graphite adopted in the UK, specifically lateral creep strain behaviour and the effect of irradiation induced creep strain on material properties, e.g. CTE and Poisson’s Ratio. A recent review of irradiation induced creep (Davies & Bradford, 2004) included an anlaysis of the available OC-series data (Mobasheran, 1990) and led to a request to ORNL for an examination of the original OC-Series dataset. An initial search of the ORNL archive revealed additional data from the OC-Series experiment including previously unknown irradiation annealing experiments. This report presents a re-analysis of the available data from the OC-Series archive.

  10. Experimental Plan for EDF Energy Creep Rabbit Graphite Irradiations- Rev. 2 (replaces Rev. 0 ORNL/TM/2013/49).

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL

    2014-07-01

    The experimental results obtained here will assist in the development and validation of future models of irradiation induced creep of graphite by providing the following data: Inert creep stain data from low to lifetime AGR fluence Inert creep-property data (especially CTE) from low to lifetime AGR fluence Effect of oxidation on creep modulus (by indirect comparison with experiment 1 and direct comparison with experiment 3 NB. Experiment 1 and 3 are not covered here) Data to develop a mechanistic understanding, including oAppropriate creep modulus (including pinning and high dose effects on structure) oInvestigation of CTE-creep strain behavior under inert conditions oInformation on the effect of applied stress/creep strain on crystallite orientation (requires XRD) oEffect of creep strain on micro-porosity (requires tomography & microscopy) This document describes the experimental work planned to meet the requirements of project technical specification [1] and EDF Energy requests for additional Pre-IE work. The PIE work is described in detail in this revision (Section 8 and 9).

  11. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or

  12. ORNL irradiation creep facility

    International Nuclear Information System (INIS)

    Reiley, T.C.; Auble, R.L.; Beckers, R.M.; Bloom, E.E.; Duncan, M.G.; Saltmarsh, M.J.; Shannon, R.H.

    1980-09-01

    A machine was developed at ORNL to measure the rates of elongation observed under irradiation in stressed materials. The source of radiation is a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). This choice allows experiments to be performed which simulate the effects of fast neutrons. A brief review of irradiation creep and experimental constraints associated with each measurement technique is given. Factors are presented which lead to the experimental choices made for the Irradiation Creep Facility (ICF). The ICF consists of a helium-filled chamber which houses a high-precision mechanical testing device. The specimen to be tested must be thermally stabilized with respect to the temperature fluctuations imposed by the particle beam which passes through the specimen. Electrical resistance of the specimen is the temperature control parameter chosen. Very high precision in length measurement and temperature control are required to detect the small elongation rates relevant to irradiation creep in the test periods available (approx. 1 day). The apparatus components and features required for the above are presented in some detail, along with the experimental procedures. The damage processes associated with light ions are discussed and displacement rates are calculated. Recent irradiation creep results are given, demonstrating the suitability of the apparatus for high resolution experiments. Also discussed is the suitability of the ICF for making high precision thermal creep measurements

  13. AGC 2 Irradiation Creep Strain Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  14. Uni axial tensile graphite creep capsules with continuous strain registration

    International Nuclear Information System (INIS)

    Hausen, H.; Loelgen, R.; Cundy, M.

    1977-01-01

    Two irradiation devices are described for the in pile measurement of tensile irradiation creep of graphite. In each machine a single sample is maintained under a controlled load by a pneumatic bellows system. Irradiation creep is measured relative to unstressed reference shells which surround the stressed sample. This differential strain is detected by linear displacement transducers, and recorded automatically by the out of pile installation. Irradiation temperatures are in the 800 to 1100 0 C range, and the stresses up to 60% of the U.T.S. One machine has been specifically designed for a flux change experiment, other irradiation parameters remaining fixed. Temperature control is achieved through varying gas mixtures in control gas gaps. The paper details the design principles of the machines and gives an account of the cold and hot commissioning tests, with particular reference to the accuracy of the in pile measuring system. Finally, the early irradiation experience is evaluated

  15. Measuring irradiation creep

    International Nuclear Information System (INIS)

    Pelah, I.

    1981-03-01

    Simulation of fusion-neutron induced damage by beams of light ions is discussed. It is suggested that accelerated creep measurements to determine ''end of life'' of materials may be done by the application of thermal treatment and thermal creep measurements. (author)

  16. Review of recent irradiation-creep results

    International Nuclear Information System (INIS)

    Coghlan, W.A.

    1982-05-01

    Materials deform faster under stress in the presence of irradiation by a process known as irradiation creep. This phenomenon is important to reactor design and has been the subject of a large number of experimental and theoretical investigations. The purpose of this work is to review the recent experimental results to obtain a summary of these results and to determine those research areas that require additional information. The investigations have been classified into four subgroups based on the different experimental methods used. These four are: (1) irradiation creep using stress relaxation methods, (2) creep measurements using pressurized tubes, (3) irradiation creep from constant applied load, and (4) irradiation creep experiments using accelerated particles. The similarity and the differences of the results from these methods are discussed and a summary of important results and suggested areas for research is presented. In brief, the important results relate to the dependence of creep on swelling, temperature, stress state and alloying additions. In each of these areas new results have been presented and new questions have arisen which require further research to answer. 65 references

  17. Modelling property changes in graphite irradiated at changing irradiation temperature

    CSIR Research Space (South Africa)

    Kok, S

    2011-01-01

    Full Text Available A new method is proposed to predict the irradiation induced property changes in nuclear; graphite, including the effect of a change in irradiation temperature. The currently used method; to account for changes in irradiation temperature, the scaled...

  18. Changes in creep of polymethylmetacrylate after irradiation

    International Nuclear Information System (INIS)

    Peschanskaya, N.N.; Smolyanskij, A.S.; Suvorova, V.Yu.

    1992-01-01

    A study was made on PMMA, irradiated by different doses of 60 Co γ-radiation in vacuum under creep during compression. It is shown that occurence of tendency to failure at +20 degC is observed at doses of D > 100 kGy (> 10 Mrad), whereas sufficient decrease of deformation before failure takes place at D > 350 kGy. Peculiarities of behaviour of irradiated and nonirradiated PMMA under compression and tension were correlated. It is noted that critical irradiation doses may differ sufficiently for different loading conditions, deformation and longevity characteristics

  19. Irradiation creep due to SIPA-induced growth

    International Nuclear Information System (INIS)

    Woo, C.H.

    1980-01-01

    An additional contribution to irradiation creep resulting from the stress-induced preferred adsorption (SIPA) effect is described - SIPA-induced growth (SIG). The mechanism of SIG is discussed and an expression for its contribution to irradiation creep developed. It is shown that SIG is very significant in comparison with SIPA. Enhancement of creep by swelling may also occur. (U.K.)

  20. Irradiation creep in zirconium single crystals

    International Nuclear Information System (INIS)

    MacEwen, S.R.; Fidleris, V.

    1976-07-01

    Two identical single crystals of crystal bar zirconium have been creep tested in reactor. Both specimens were preirradiated at low stress to a dose of about 4 x 10 23 n/m 2 (E > 1 MeV), and were then loaded to 25 MPa. The first specimen was loaded with reactor at full power, the second during a shutdown. The loading strain for both crystals was more than an order of magnitude smaller than that observed when an identical unirradiated crystal was loaded to the same stress. Both crystals exhibited periods of primary creep, after which their creep rates reached nearly constant values when the reactor was at power. During shutdowns the creep rates decreased rapidly with time. Electron microscopy revealed that the irradiation damage consisted of prismatic dislocation loops, approximately 13.5 nm in diameter. Cleared channels, identified as lying on (1010) planes, were also observed. The results are discussed in terms of the current theories for flux enhanced creep in the light of the microstructures observed. (author)

  1. Irradiation creep, stress relaxation and a mechanical equation of state

    International Nuclear Information System (INIS)

    Foster, J.P.

    1976-01-01

    Irradiation creep and stress relaxation data are available from the United Kingdom for 20 percent CW M316, 20 percent CW FV 548 and FHT PE16 using pure torsion in the absence of swelling at 300 0 C. Irradiation creep models were used to calculate the relaxation and permanent deflection of the stress relaxation tests. Two relationships between irradiation creep and stress relaxation were assessed by comparing the measured and calculated stress relaxation and permanent deflection. The results show that for M316 and FV548, the stress relaxation and deflection may be calculated using irradiation creep models when the stress rate term arising from the irradiation creep model is set equal to zero. In the case of PE16, the inability to calculate the stress relaxation and permanent deflection from the irradiation creep data was attributed to differences in creep behavior arising from lot-to-lot variations in alloying elements and impurity content. A modification of the FV548 and PE16 irradiation creep coefficients was necessary in order to calculate the stress relaxation and deflection. The modifications in FV548 and PE16 irradiation creep properties reduces the large variation in the transient or incubation parameter predicted by irradiation creep tests for M316, FV548 and PE16

  2. Irradiation creep under 60 MeV alpha irradiation

    International Nuclear Information System (INIS)

    Reiley, T.C.; Shannon, R.H.; Auble, R.L.

    1980-01-01

    Accelerator-produced charged-particle beams have advantages over neutron irradiation for studying radiation effects in materials, the primary advantage being the ability to control precisely the experimental conditions and improve the accuracy in measuring effects of the irradiation. An apparatus has recently been built at ORNL to exploit this advantage in studying irradiation creep. These experiments employ a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). The experimental approach and capabilities of the apparatus are described. The damage cross section, including events associated with inelastic scattering and nuclear reactions, is estimated. The amount of helium that is introduced during the experiments through inelastic processes and through backscattering is reported. Based on the damage rate, the damage processes and the helium-to-dpa ratio, the degree to which fast reactor and fusion reactor conditions may be simulated is discussed. Recent experimental results on the irradiation creep of type 316 stainless steel are presented, and are compared to light ion results obtained elsewhere. These results include the stress and temperature dependence of the formation rate under irradiation. The results are discussed in relation to various irradiation creep mechanisms and to damage microstructure as it evolves during these experiments. (orig.)

  3. Irradiation creep induced anisotropy in a/2 dislocation populations

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1984-05-01

    The contribution of anisotropy in Burgers vector distribution to irradiation creep behavior has been largely ignored in irradiation creep models. However, findings on Frank loops suggest that it may be very important. Procedures are defined to identify the orientations of a/2 Burgers vectors for dislocations in face-centered cubic crystals. By means of these procedures the anisotropy in Burgers vector populations was determined for three Nimonic PE16 pressurized tube specimens irradiated under stress. Considerable anisotropy in Burgers vector population develops during irradiation creep. It is inferred that dislocation motion during irradiation creep is restricted primarily to a climb of a/2 dislocations on 100 planes. Effect of these results on irradiation creep modeling and deformation induced irradiation growth is considered

  4. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  5. A constitutive equation of irradiation creep and swelling under neutron irradiation

    International Nuclear Information System (INIS)

    Murakami, Sumio; Mizuno, Mamoru; Okamoto, Toshiaki.

    1990-01-01

    A constitutive equation of irradiation creep for irradiated materials applicable to structural analyses in a multiaxial state of stress was developed. After reviewing microscopic mechanisms of irradiation creep and swelling, the relevant theories proposed from the view point of metallurgical physics and their applicability were discussed first. Then a constitutive model was developed by assuming that irradiation creep can be decomposed into irradiation-enhanced creep and irradiation-induced creep. By taking account of the SIPA (Stress Induced Preferential Absorption) mechanism, the irradiation-induced creep was represented by an isotropic tensor function of order one and zero with respect to stress, which is, at the same time, the function of neutron flux and neutron fluence. The volumetric part of the irradiation-induced creep was identified with swelling. The irradiation-enhanced creep was described by modifying Kachanov-Rabotnov creep damage theory by incorporating the effect of irradiation. Finally the irradiation creep and swelling of type 316 stainless steel at elevated temperatures were predicted by the proposed constitutive equation, and the numerical results were compared with the corresponding experimental results. (author)

  6. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  7. Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2017-10-01

    Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

  8. Property changes in graphite irradiated at changing irradiation temperature

    International Nuclear Information System (INIS)

    Price, R.J.; Haag, G.

    1979-07-01

    Design data for irradiated graphite are usually presented as families of isothermal curves showing the change in physical property as a function of fast neutron fluence. In this report, procedures for combining isothermal curves to predict behavior under changing irradiation temperatures are compared with experimental data on irradiation-induced changes in dimensions, Young's modulus, thermal conductivity, and thermal expansivity. The suggested procedure fits the data quite well and is physically realistic

  9. Irradiation creep and growth of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Lansiart, S.; Darchis, L.; Pelchat, J.

    1990-01-01

    The influence of temperature and fast neutron flux on irradiation creep and growth of stress relieved zircaloy-4 pressurized tubes has been derived from experimental irradiations in NaK, performed up to 2.5 10 25 n.m -2 in the temperature range [280, 350] 0 C. A significant influence of temperature on axial growth has been observed: at 280 0 C the elongation can no longer be expressed as a linear function of fluence as for the 350 0 C irradiation temperature; diametral growth, on the other hand, always appears negligible. Irradiation creep obviously depends on temperature too; the diametral strain (including thermal part) has been modelled as a sum of primary and secondary terms, the former being independent of fluence. For the tubing considered it is observed that the ranking of the different batches, with respect to diametral creep resistance, is the same before and under irradiation. Concerning axial creep strain the stress relieved material behaves as does an isotropic tube. This is not the case of recrystallized zircaloy-4 F, which shows a non negligible axial deformation, related to the diametral creep one, even though this diametral irradiation creep strain is strongly reduced comparatively to that of the stress relieved material. The comparison of the two materials growth rates is more complex since their dependence on temperature and flux differs

  10. Ion irradiated graphite exposed to fusion-relevant deuterium plasma

    International Nuclear Information System (INIS)

    Deslandes, Alec; Guenette, Mathew C.; Corr, Cormac S.; Karatchevtseva, Inna; Thomsen, Lars; Ionescu, Mihail; Lumpkin, Gregory R.; Riley, Daniel P.

    2014-01-01

    Graphite samples were irradiated with 5 MeV carbon ions to simulate the damage caused by collision cascades from neutron irradiation in a fusion environment. The ion irradiated graphite samples were then exposed to a deuterium plasma in the linear plasma device, MAGPIE, for a total ion fluence of ∼1 × 10 24 ions m −2 . Raman and near edge X-ray absorption fine structure (NEXAFS) spectroscopy were used to characterize modifications to the graphitic structure. Ion irradiation was observed to decrease the graphitic content and induce disorder in the graphite. Subsequent plasma exposure decreased the graphitic content further. Structural and surface chemistry changes were observed to be greatest for the sample irradiated with the greatest fluence of MeV ions. D retention was measured using elastic recoil detection analysis and showed that ion irradiation increased the amount of retained deuterium in graphite by a factor of four

  11. Irradiation creep of the mixed oxide UPuO2

    International Nuclear Information System (INIS)

    Combette, Patrick; Milet, Claude

    1976-01-01

    The irradiation creep under compression of the mixed oxide UO 2 -PuO 2 was studied up to fission yields of 6x10 13 fcm -3 s -1 , under stresses -2 , in the temperature range 700-900 deg C. The creep rate is proportional to the applied stress and fission yield, athermal in the studied temperature range and non-dependent of burnup (up to 30000MWjt -1 ). In a sample under compression, swelling is observed due to the formation of fission products during the irradiation and the swelling rate is of the same order that in a cladded fuel element [fr

  12. Neutron irradiation effects on graphite cloth

    International Nuclear Information System (INIS)

    Gray, W.J.

    1976-01-01

    A series of cloth and fiber samples has been irradiated to fluences of 3.5, 7.3, and 10 x 10 21 cm -2 EFF* at 470 0 C. Data from the first set of samples show large shrinkages relative to that found for typical nuclear graphites. Nevertheless, all but one of the 2-dimensional cloths were unchanged except for the shrinkage. The 3-dimensional cloths, on the other hand, have deteriorated apparently because these types of weaves are less able to accommodate the large axial fiber shrinkages

  13. Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Murty, Korukonda [North Carolina State Univ., Raleigh, NC (United States); Burchell, Timothy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-06-02

    Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

  14. Development of integrated waste management options for irradiated graphite

    Directory of Open Access Journals (Sweden)

    Alan Wareing

    2017-08-01

    Full Text Available The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  15. Development of integrated waste management options for irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, Alan; Abrahamsen-Mills, Liam; Fowler, Linda; Jarvis, Richard; Banford, Anthony William [National Nuclear Laboratory, Warrington (United Kingdom); Grave, Michael [Doosan Babcock, Gateshead (United Kingdom); Metcalfe, Martin [National Nuclear Laboratory, Gloucestershire (United Kingdom); Norris, Simon [Radioactive Waste Management Limited, Oxon (United Kingdom)

    2017-08-15

    The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  16. Mechanical modeling of creep, swelling and damage under irradiation for polycrystalline metals

    International Nuclear Information System (INIS)

    Murakami, S.; Mizuno, M.; Okamoto, T.

    1991-01-01

    A constitutive equation of creep, swelling and damage under irradiation for polycrystalline metals applicable to structural analyses in multiaxial state of stress is developed. After reviewing microscopic mechanisms of irradiation creep and swelling, the relevant theories proposed so far from the view point of metallurgical physics and their applicability are discussed first. Then a constitutive model is developed by assuming that creep under irradiation can be decomposed into irradiation-affected thermal creep and irradiation-induced creep. By taking account of the Stress-Induced Preferential Absorption (SIPA) mechanism, the irradiation-induced creep is represented by an isotropic tensor function of order one and zero with respect to stress, which is, at the same time, the function of neutron flux and neutron fluence. The volumetric part of the irradiation-induced creep is identified with swelling. The irradiation-affected thermal creep is described by modifying Kachanov-Rabotnov theory for stress-controlled creep and creep damage by incorporating the effect of irradiation. Finally irradiation creep and swelling of 20% cold-worked type 316 stainless steel at elevated temperature are predicted by the proposed constitutive equations, and the numerical results are compared with the corresponding experimental results. (orig.)

  17. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    International Nuclear Information System (INIS)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-01-01

    The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.

  18. The behavior of interstitials in irradiated graphite

    International Nuclear Information System (INIS)

    Pedraza, D.F.

    1991-01-01

    A computer model is developed to simulate the behavior of self-interstitials with particular attention to clustering. Owing to the layer structure of graphite, atomistic simulations can be performed using a large parallelepipedic supercell containing a few layers. In particular, interstitial clustering is studied here using a supercell that contains two basal planes only. Frenkel pairs are randomly produced. Interstitials are placed at sites between the crystal planes while vacancies are distributed in the two crystal planes. The size of the computational cell is 20000 atoms and periodic boundary conditions are used in two dimensions. Vacancies are assumed immobile whereas interstitials are given a certain mobility. Two point defect sinks are considered, direct recombination of Frenkel pairs and interstitial clusters. The clusters are assumed to be mobile up to a certain size where they are presumed to become loop nuclei. Clusters can shrink by emission of singly bonded interstitials or by recombination of a peripheral interstitial with a neighboring vacancy. The conditions under which interstitial clustering occurs are reported. It is shown that when clustering occurs the cluster size population gradually shifts towards the largest size cluster. The implications of the present results for irradiation growth and irradiation-induced amorphization are discussed

  19. Swelling and irradiation creep of neutron irradiated 316Ti and 15-15Ti steels

    International Nuclear Information System (INIS)

    Maillard, A.; Touron, H.; Seran, J.L.; Chalony, A.

    1992-01-01

    The global behavior, the swelling and irradiation creep resistances of cold worked 316Ti and 15-15Ti, two variants of austenitic steels in use as core component materials of the French fast reactors, are compared. The 15-15Ti leads to a significant improvement due to an increase in the incubation dose swelling. The same phenomena observed on 316Ti are found on 15-15Ti. All species without fuel like samples, wrappers or empty clad swell and creep less than fuel pin cladding irradiated in the same conditions. To explain the swelling difference, as for 316Ti, thermal gradient is also invoked but the irradiation creep difference is not yet clearly understood. To predict the behavior of clads it is indispensable to study the species themselves and to use specific rules. All results confirm the good behavior of 15-15Ti, the best behavior being obtained with the 1% Si doped version irradiated up to 115 dpa

  20. AGC-2 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

  1. Variation of the properties of siliconized graphite during neutron irradiation

    International Nuclear Information System (INIS)

    Virgil'ev, Y.S.; Chugunova, T.K.; Pikulik, R.G.

    1986-01-01

    The authors evaluate the radiation-induced property changes in siliconized graphite of the industrial grades SG-P and SG-M. The authors simultaneously tested the reference (control) specimens of graphite that are used as the base for obtaining the SG-M siliconized graphite by impregnating with silicon. The suggested scheme (model) atributes the dimensional changes of the siliconized graphite specimens to the effect of the quantitative ratio of the carbide phase and carbon under different conditions of irradiation. If silicon is insufficient for the formation of a dense skeleton, graphite plays a devisive role, and it may be assumed that at an irradiation temperature greater than 600 K, the material shrinks. The presence of isolated carbide inclusions also affects the physicomechanical properties (including the anitfriction properties)

  2. Data acquisition system for light-ion irradiation creep experiment

    International Nuclear Information System (INIS)

    Hendrick, P.L.; Whitaker, T.J.

    1979-07-01

    Software was developed for a PDP11V/03-based data acquisition system to support the Light-Ion Irradiation Creep Experiment installed at the University of Washington Tandem Van de Graaff Accelerator. The software consists of a real-time data acquisition and storage program, DAC04, written in assembly language. This program provides for the acquisition of up to 30 chennels at 100 Hz, data averaging before storage on disk, alarming, data table display, and automatic disk switching. All analog data are acquired via an analog-to-digital converter subsystem having a resolution of 14 bits, a maximum throughput of 20 kHz, and an overall system accuracy of +-0.01%. These specifications are considered essential for the long-term measurement of irradiation creep strains and temperatures during the light-ion bombardment of irradiation creep specimens. The software package developed also contains a collection of FORTRAN programs designed to monitor a test while in progress. These programs use the foreground/background feature of the RT-11 operating system. The background programs provide a variety of services. The program, GRAFTR, allows transient data (i.e., prior to averaging) to be graphed at the graphics terminal. The program, GRAFAV, allows averaged data to be read from disk and displayed graphically at the terminal. The program, TYPAV, reads averaged data from disk and displays it at the terminal in tabular form. Other programs allow text messages to be written to disk, read from disk, and allow access to DAC04 initialization data. 5 figures, 18 tables

  3. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  4. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  5. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marcu, A., E-mail: aurelian.marcu@inflpr.ro [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Avotina, L. [Institute of Chemical Physics, University of Latvia, Kronvalda 4, LV 1010 Riga (Latvia); Porosnicu, C. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Marin, A. [Ilie Murgulescu” Institute of Physical Chemistry, 202 Splaiul Independentei 060021, Bucharest (Romania); Grigorescu, C.E.A. [National Institute R& D for Optoelectronics INOE 2000, 077125 Bucharest (Romania); Ursescu, D. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Lungu, M. [National Institute of Materials Physics Atomistilor Str., 105 bis, 077125, Magurele (Romania); Demitri, N. [Hard X-ray Beamline and Structural Biology, Elettra-Sincrotrone Trieste, Strada Statale 14 - km 163,5 in AREA Science Park, 34149 Basovizza TS Italy (Italy); Lungu, C.P. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania)

    2015-11-15

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp{sup 3}% increase at tens of nm below the surface. • sp{sup 3}% is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp{sup 3} bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp{sup 3} percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  6. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    International Nuclear Information System (INIS)

    Marcu, A.; Avotina, L.; Porosnicu, C.; Marin, A.; Grigorescu, C.E.A.; Ursescu, D.; Lungu, M.; Demitri, N.; Lungu, C.P.

    2015-01-01

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp 3 % increase at tens of nm below the surface. • sp 3 % is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp 3 bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp 3 percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  7. Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1978-01-01

    This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

  8. Irradiation creep lifetime analysis on first wall structure materials for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

    2017-05-15

    Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

  9. Reflection and photoemission studies of neutron-irradiated graphite

    International Nuclear Information System (INIS)

    Fukutani, Hirohito; Yamada, Akio; Yagi, Kazutoshi; Ooe, Satoshi; Higashiyama, Kazuyuki; Kato, Hiroo; Iwata, Tadao.

    1990-01-01

    Neutron-irradiated graphites were studied by reflectivity and photoemission (UPS, ARUPS, XPS) measurements. The π-band reflectivity peak of graphite, located at 5 eV, changed significantly and a small absorption band ascribed to vacancies produced by neutron bombardment was found to grow around 3 eV. Modification of the valence band by neutron irradiation was studied by ARUPS. The π-valence band shifts to lower binding energy towards the Fermi level and its band width becomes smaller. These results were also confirmed by the optical joint density of states obtained from K-K analysis of the reflectivity. (author)

  10. Final report on graphite irradiation test OG-3

    International Nuclear Information System (INIS)

    Price, R.J.; Beavan, L.A.

    1977-01-01

    The results of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on graphite specimens irradiated in capsule OG-3 are presented. The graphite grades investigated included near-isotropic H-451 (three different preproduction lots), TS-1240, and SO818; needle coke H-327; and European coal tar pitch coke grades P 3 JHA 2 N, P 3 JHAN, and ASI2-500. Data were obtained in the temperature range 823 0 K to 1673 0 K. The peak fast neutron fluence in the experiment was 3 x 10 25 n/m 3 (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10 25 n/m 2 on some H-451 specimens and 6 x 10 25 n/m 2 on some TS-1240 specimens. Irradiation-induced dimensional changes on H-451 graphite differed slightly from earlier predictions. For an irradiation temperature of about 1225 0 K, axial shrinkage rates at high fluences were somewhat higher than predicted, and the fluence at which radial expansion started (about 9 x 10 25 n/m 2 at 1275 0 K) was lower. TS-1240 graphite underwent smaller dimensional changes than H-451 graphite, while limited data on SO818 and ASI2-500 graphites showed similar behavior to H-451. P 3 JHAN and P 3 JHA 2 N graphites displayed anisotropic behavior with rapid axial shrinkage. Comparison of dimensional changes between specimens from three logs of H-451 and of TS-1240 graphites showed no significant log-to-log variations for H-451, and small but significant log-to-log variations for TS-1240. The thermal expansivity of the near-isotropic graphites irradiated at 865-1045 0 K first increased by 5 percent to 10 percent and then decreased. At higher irradiation temperatures the thermal expansivity decreased by up to 50 percent. Changes in thermal conductivity were consistent with previously established curves. Specimens which were successively irradiated at two different temperatures took on the saturation conductivity for the new temperature

  11. Irradiation creep of nano-powder sintered silicon carbide at low neutron fluences

    International Nuclear Information System (INIS)

    Koyanagi, T.; Shimoda, K.; Kondo, S.; Hinoki, T.; Ozawa, K.; Katoh, Y.

    2014-01-01

    The irradiation creep behavior of nano-powder sintered silicon carbide was investigated using the bend stress relaxation method under neutron irradiation up to 1.9 dpa. The creep deformation was observed at all temperatures ranging from 380 to 1180°C mainly from the irradiation creep but with the increasing contributions from the thermal creep at higher temperatures. The apparent stress exponent of the irradiation creep slightly exceeded unity, and instantaneous creep coefficient at 380 to 790°C was estimated to be ∼1 × 10 -5 [MPa -1 dpa -1 ] at ∼0.1 dpa and 1 × 10 -7 to 1 × 10 -6 [MPa -1 dpa -1 ] at ∼1 dpa. The irradiation creep strain appeared greater than that for the high purity SiC. Microstructural observation and data analysis indicated that the grain-boundary sliding associated with the secondary phases contributes to the irradiation creep at 380–790°C to 0.01–0.11 dpa. (author)

  12. Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

    Science.gov (United States)

    Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

    2009-12-01

    Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

  13. A constitutive equation of creep, swelling and damage under neutron irradiation applicable to multiaxial and variable states of stress

    International Nuclear Information System (INIS)

    Murakami, Sumio; Mizuno, Mamoru.

    1992-01-01

    A constitutive equation of creep, swelling and damage under neutron irradiation applicable to multiaxial non-steady states of stress is developed. In the formulation of the constitutive equation, the creep under irradiation was divided into irradiation-affected thermal creep and irradiation-induced creep. Then the irradiation-affected thermal creep was formulated by extending the creep-hardening surface model to include the effects of neutron-irradiation and material damage. The Bailey-Norton creep law and Kachanov-Rabotnov creep-damage theory were employed. The effect of irradiation on thermal creep was described by expressing the material functions of the constitutive equation as functions of neutron flux φ and neutron fluence Φ. The constitutive equation of irradiation-induced creep was formulated by taking account of SIPA and CCG mechanisms and by representing the creep rate as a function of stress of order zero and one. Creep of 316 stainless steel under various conditions of irradiation and variable stress was analyzed in order to elucidate the validity and the utility of the proposed constitutive equation. (author)

  14. Irradiation creep of solution annealed and cold worked 316 stainless steel

    International Nuclear Information System (INIS)

    Boutard, J.L.; Carteret, Y.; Cauvin, R.; Maillard, A.; Guerin, Y.

    1983-01-01

    Irradiation creep strains obtained in-pile on S.A. and C.W. 316 show a linear creep-swelling correlation, the slope of which is rather insensitive to chemical composition and elements in solid solution. The variation of SIPA component resulting only from the evolution of dislocation density and void growth cannot explain such an empirical correlation. The I-creep term has, on the other hand, the right temperature dependence and order of magnitude. (author)

  15. A SIPA-based theory of irradiation creep in the low swelling rate regime

    International Nuclear Information System (INIS)

    Garner, F.A.; Woo, C.H.

    1991-11-01

    A model is presented which describes the major facets of the relationships between irradiation creep, void swelling and applied stress. The increasing degree of anisotropy in distribution of dislocation Burger's vectors with stress level plays a major role in this model. Although bcc metals are known to creep and swell at lower rates than fcc metals, it is predicted that the creep-swelling coupling coefficient is actually larger

  16. Stress state dependence of transient irradiation creep in 20% cold worked 316 stainless steel

    International Nuclear Information System (INIS)

    Foster, J.P.; Gilbert, E.R.

    1998-01-01

    Irradiation creep tests were performed in fast reactors using the stress states of uniaxial tension, biaxial tension, bending and torsion. In order to compare the saturated transient strain irradiation creep component, the test data were converted to equivalent strain and equivalent stress. The saturated transient irradiation creep component was observed to depend on the stress state. The highest value was exhibited by the uniaxial tension stress state, and the lowest by the torsion stress state. The biaxial tension and bending stress state transient component values were intermediate. This behavior appears to be related to the dislocation or microscopic substructure resulting from fabrication processing and the applied stress direction. (orig.)

  17. Erosion of pyrolytic graphite and Ti-doped graphite due to high flux irradiation

    International Nuclear Information System (INIS)

    Ohtsuka, Yusuke; Ohashi, Junpei; Ueda, Yoshio; Isobe, Michiro; Nishikawa, Masahiro

    1997-01-01

    The erosion of pyrolytic graphite and titanium doped graphite RG-Ti above 1,780 K was investigated by 5 keV Ar beam irradiation with the flux from 4x10 19 to 1x10 21 m -2 ·s -1 . The total erosion yields were significantly reduced with the flux. This reduction would be attributed to the reduction of RES (radiation enhanced sublimation) yield, which was observed in the case of isotropic graphite with the flux dependence of RES yield of φ -0.26 (φ: flux) obtained in our previous work. The yield of pyrolytic graphite was roughly 30% higher than that of isotropic graphite below the flux of 10 20 m -2 ·s -1 whereas each yield approached to very close value at the highest flux of 1x10 21 m -2 ·s -1 . This result indicated that the effect of graphite structure on the RES yield, which was apparent in the low flux region, would disappear in the high flux region probably due to the disordering of crystal structure. In the case of irradiation to RG-Ti at 1,780 K, the surface undulations evolved with a mean height of about 3 μm at 1.2x10 20 m -2 ·s -1 , while at higher flux of 8.0x10 20 m -2 ·s -1 they were unrecognizable. These phenomena can be explained by the reduction of RES of graphite parts excluding TiC grains. (author)

  18. On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

  19. Technique for measuring irradiation creep in polycrystalline SiC fibers

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Hamilton, M.L.; Jones, R.H. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    A bend stress relaxation (BSR) test has been designed to examine irradiation enhanced creep in polycrystalline SiC fibers being considered for fiber reinforcement in SiC/SiC composite. Thermal creep results on Nicalon-CG and Hi-Nicalon were shown to be consistent with previously published data with Hi-Nicalon showing about a 100{degrees}C improvement in creep resistance. Preliminary data was also obtained on Nicalon-S that demonstrated that its creep resistance is greater than that of Hi-Nicalon.

  20. Oxidation and sublimation of porous graphite during fiber laser irradiation

    Science.gov (United States)

    Phillips, Grady T.; Bauer, William A.; Gonzales, Ashley E.; Herr, Nicholas C.; Perram, Glen P.

    2017-02-01

    Porous graphite plates, cylinders and cones with densities of 1.55-1.82 g/cm3 were irradiated by a 10 kW fiber laser at 0.075 -3.525 kW/cm2 for 120 s to study mass removal and crater formation. Surface temperatures reached steady state values as high as 3767 K. The total decrease in sample mass ranged from 0.06 to 6.29 g, with crater volumes of 0.52 - 838 mm3, and penetration times for 12.7 mm thick plates as short as 38 s. Minor contaminants in the graphite samples produced calcium and iron oxide to be re-deposited on the graphite surface. Significantly increased porosity of the sample is observed even outside of the laser-irradiated region. Total mass removed increases with deposited laser energy at a rate of 4.83 g/MJ for medium extruded graphite with an apparent threshold of 0.15 MJ. Visible emission spectroscopy reveals C2 Swan and CN red, CN violet bands and Li, Na, and K 2P3/2,1/2 - 2S1/2 doublets. The reacting boundary layer is observed using a mid-wave imaging Fourier transform spectrometer (IFTS) at 2 cm-1 spectral resolution, 0.5 mm/pixel spatial resolution, and 0.75 Hz data cube rate. A two-layer radiative transfer model was used to determine plume temperature, CO, and CO2 concentrations from spectral signatures. The new understanding of graphite combustion and sublimation during laser irradiation is vital to the more complex behavior of carbon composites.

  1. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  2. Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

    International Nuclear Information System (INIS)

    Kalinowski, J.E.

    1979-01-01

    Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

  3. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

    2000-02-01

    This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

  4. Change in electrical resistance of irradiated nuclear graphite during compressive tests

    International Nuclear Information System (INIS)

    Eto, Motokuni

    1986-01-01

    In this paper the change in electrical resistance of neutron-irradiated nuclear graphite is measured and compared with that of the unirradiated; the deformation mechanism for irradiated graphite is discussed in relation to the resistance change and the defects which are believed to be produced during irradiation. (orig./RK)

  5. Stress relaxation analysis and irradiation creep and swelling in pressure tubes

    International Nuclear Information System (INIS)

    Beeston, J.M.; Burr, T.K.

    1979-01-01

    An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

  6. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    Ribis, J.

    2007-12-01

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  7. Bending creep in the direction perpendicular to grain during microwave irradiation

    International Nuclear Information System (INIS)

    Iida, I.

    1989-01-01

    Bending creep tests in the radial direction perpendicular to the grain were carried out on the thirteen different wood species during the microwave irradiation and during the hot-air drying. The course of moisture content of specimen during creep tests were measured at the same time. And then, relationships between the drying rate and the moisture content, or the creep deflection and the moisture content were investigated and disscussed. Results obtained are as follows : 1) The coefficients of drying rate (K 1 ) during microwave irradiation process were from values of 3.40(hr) -1 to 5.65(hr) -1 for different species. With average value of all woods, there were of 4.73(hr) -1 . Therefore, this value show a value of 5.3 times as much as these of hot-air drying. 2) Creep deflection of woods dried by the microwave heating increase remarkably from the start of the microwave irradiation. 3) Ratio ( y 30 /y m ) of creep deflection y m , in region of ∼30% moisture content, to the maximum creep deflection y m were thought the values differ from each wood species, in no relation with the applied stresses and these values have the constant in a wood. Those were estimated about 0.73 for Icho wood and about 0.44 for Buna wood, and moreover it was about 0.6 with average value for all wood species. Consequently, it was recognized that drying rate became remarkably magnitude value during microwave heating. Creep deflection on the 30% moisture content take beyond about half of the total creep deflection. Conseqently, the large creep deformation developed during the high moisture content process, and it constitute a caractaristic frature of microwave heating

  8. Graphite oxidation and damage under irradiation at high temperatures in an impure helium environment

    Science.gov (United States)

    Goodwin, Cameron S.

    The High Temperature Gas-Cooled Reactor (HTGR) is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. In order to investigate the mechanism for corrosion of graphite in HTGRs, the graphite was placed in a similar environment in order to evaluate its resistance to corrosion and oxidation. While the effects of radiation on graphite have been studied in the past, the properties of graphite are largely dependent on the coke used in manufacturing the graphite. There are no longer any of the previously studied graphite types available for use in the HTGR. There are various types of graphite being considered for different uses in the HTGR and all of these graphite types need to be analyzed to determine how radiation will affect them. Extensive characterization of samples of five different types of graphite was conducted. The irradiated samples were analyzed with electron paramagnetic resonance spectroscopy, Raman spectroscopy, x-ray diffraction, x-ray photoelectron spectroscopy and gas chromatography. The results prove a knowledge base for considering the graphite types best suited for use in HTGRs. In my dissertation work graphite samples were gamma irradiated and also irradiated in a mixed field, in order to study the effects of neutron as well as gamma irradiation. Thermal effects on the graphite were also investigated by irradiating the samples at room temperature and at 1000 °C. From the analysi of the samples in this study there is no evidence of substantial damage to the grades of graphite analyzed. This is significant in approving the use of these graphites in nuclear reactors. Should significant damage had occurred to the samples, the use of these grades of graphite would need to be reconsidered. This information can be used to further characterize other grades of nuclear graphite as they become available.

  9. Neutron irradiation effects on carbon and graphite cloths and fibers

    International Nuclear Information System (INIS)

    Gray, W.J.

    1977-08-01

    A series of cloth and fiber samples were irradiated to fluences of 3.5, 7.3, and 10 x 10 21 cm -2 at 470 0 C. Dimensional changes of the fibers in the radial direction ranged from -19% to +33% and in the axial direction from -18% to -27%, roughly ten times greater than dimensional changes found for typical nuclear graphites. Despite these large dimensional changes, all but one of the 2-dimensional cloths remained essentially unchanged in overall physical appearance. The 3-dimensional cloths, on the other hand, deteriorated apparently because these types of weaves were less able to accommodate the large axial fiber shrinkages

  10. Effects of neutron irradiation on tensile and creep properties of stainless steels

    International Nuclear Information System (INIS)

    Miyaji, Noriko; Abe, Yasuhiro; Asayama, Tai; Aoto, Kazumi; Ukai, Shigeharu

    1997-01-01

    In order to investigate the effects of neutron irradiation on the creep and tensile properties of stainless steels, post-irradiation tests were made on the specimens of FBR grade type 316 stainless steel (316FR) and type 304 stainless steel. The post-irradiation tensile tests showed that the fracture elongation of both 316FR and type 304 stainless steel decreased and the 0.2% proof strength increased by irradiation. These phenomena are related to the point defect accumulation due to neutron irradiation. The post-irradiation creep test of 316FR demonstrated that the time to rupture decreased to between 1/3 and 1/5 of the unirradiated one, and this reduction is smaller than that of type 304 stainless steels under the same irradiation and test conditions. The creep property degradation of type 304 stainless steel due to the irradiation is caused by accumulation of helium bubbles at the grain boundaries. As for 316FR, it is considered that beyond the neutron exposure level of 0.3dpa a growth of phosphide caused a decrease in solution hardening and accumulation of helium bubbles at the grain boundaries. It is concluded that the reduction ratio of time to rupture for both 316FR and type 304 stainless steels after irradiation became larger than 1/30, which is the lower limit of the reduction ratio for the 'Monju' FBR. (author)

  11. Effects of prior stress history on the irradiation creep of 20% cold-worked AISI 316 stainless steel

    International Nuclear Information System (INIS)

    Chin, B.A.; Straalsund, J.L.; Wire, G.L.

    1979-01-01

    The following conclusions resulted from this study: An in-reactor transient component of creep is found to occur whenever the stress level is increased. The transient is principally a thermal process, short in duration, and only weakly dependent on flux. The observed irradiation component of in-reactor creep is independent of prior stress history. Microstructural development during irradiation is influenced predominantly by the irradiation flux and temperature variables, and only to a minor extent by the irradiation stress history. (Auth.)

  12. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

    The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

  13. The study of creep in stainless steel irradiated with fast neutron and alpha particles

    International Nuclear Information System (INIS)

    Correa, D.A.C.

    1985-01-01

    The objective of the present work is to study the creep behavior of the 316 type stainless steel 50% cold worked in different conditions of temperature and applied stress, after neutron radiation and Alfa particles implantation. For this experiment, non-irradiated samples, samples irradiated in the research reactor IEA-R1 with fast neutron (E≥ MeV) up to a fluence of 8.6.10 17 n/cm 2 , and samples implanted with Alfa particles in the cyclotron CV-28 with Helium concentrations of 5 and 26 appm, were creep tested with applied stresses of the 200-300 MPa at temperatures between 650 0 C and 700 0 C. The deformation versus time curves were plotted and it was observed tha the second stage is not well defined, with the creep rate increasing continuously until the occurrence of failure of the material. The study of the effect of increase from 200 MPa to 300 MPa at the same temperature was performed. It can be concluded that this increase produces an approximately 70% reductions in the fracture time of the material, with practically no influence in the total deformation. Samples were tested at different temperatures (650, 675 and 700 0 C) at a same applied stress (200 MPa). It has been observed that a temperature of 50 0 C produces 98,9% of reduction in the fracture time and almost doubles the total deformation. On neutron irradiated samples, creep tests were performed at the same temperature and stress of the non irradiated samples. Comparing the results obtained a tendency of embrittlement due to the neutron irradiation can be observed; no remarkable structure changes were detected due to small fast neutron. Microstructural and metalographic observations were performed before and after each creep test. (author) [pt

  14. Environmental effects on irradiation creep behavior of highly purified V-4Cr-4Ti alloys (NIFS-Heats) irradiated by neutrons

    OpenAIRE

    FUKUMOTO, K; MARUI, M; MATSUI, H; NAGASAKA, T; MUROGA, T; LI, M; HOELZER, D.T.; ZINKLE, S.J.

    2009-01-01

    In order to investigate the effect of the environment on the irradiation creep properties ofhighly purified V-4Cr-4Ti alloys, neutron irradiation experiments with sodium-enclosed irradiationcapsules in Joyo and lithium-enclosed irradiation capsules in HFIR-17J were carried out usingpressurized creep tubes (PCTs).It was found that the creep strain rate exhibited a linear relationship with the effective stressup to 150 MPa at 458°C and 598°C in the Joyo irradiation experiments. For HFIR-17J irr...

  15. Irradiation-induced creep and microstructural development in precipitation-hardened nickel-aluminium alloys

    International Nuclear Information System (INIS)

    Ansari, I.

    1985-04-01

    Irradiation-induced creep in solid-solution Ni-8.5 at% AL and precipitation-hardened Ni-13.1 at% Al alloys was studied by bombarding miniaturized specimens with 6.2 MeV protons at 300 0 C under different tensile stresses. After irradiation transmission electron microscopic (TEM) investigations were made to observe the precipitate structure under irradiation for different experimental parameters. Moreover, the irradiation-induced changes in precipitate structure and changes of Al-concentrations in the matrix in Ni-13.1 at% Al alloys were studied by electrical resistivity measurements during irradiation. For comparison, the electrical resistivity of unirradiated specimens was also measured after thermal aging for different times. For correlation, TEM analysis was performed on irradiated and unirradiated aged specimens. Tensile tests on annealed and aged Ni-Al alloys were also done at various temperatures. (orig./RK)

  16. Study on 14C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the 14 C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of 14 C in fuel elements and graphite spheres of HTR-10, the production mechanism of the 14 C in graphite spheres was studied. The production sources of the 14 C in graphite spheres and fuel elements were summarized, the amount of 14 C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the 14 C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  17. Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

    International Nuclear Information System (INIS)

    Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

    2001-01-01

    It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

  18. AGC-3 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

    This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to

  19. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

  20. The reaction of unirradiated and irradiated nuclear graphites with water vapor in helium

    International Nuclear Information System (INIS)

    Imai, Hisashi; Nomura, Shinzo; Kurosawa, Takeshi; Fujii, Kimio; Sasaki, Yasuichi

    1980-10-01

    Nuclear graphites more than 10 brands were oxidized with water vapor in helium and then some selected graphites were irradiated with fast neutron in the Japan Materials Testing Reactor to clarify the effect of radiation damage of graphite on their reaction behaviors. The reaction was carried out under a well defined condition in the temperature range 800 -- 1000 0 C at concentrations of water vapor 0.38 -- 1.30 volume percent in helium flow of total pressure of 1 atm. The chemical reactivity of graphite irradiated at 1000 +- 50 0 C increased linearly with neutron fluence until irradiation of 3.2 x 10 21 n/cm 2 . The activation energy for the reaction was found to decrease with neutron fluence for almost all the graphites, except for a few ones. The order of reaction increased from 0.5 for the unirradiated graphite to 1.0 for the graphite irradiated up to 6.0 x 10 20 n/cm 2 . Experiment was also performed to study a superposed effect between the influence of radiation damage of graphite and the catalytic action of barium on the reaction rate, as well as the effect of catalyser of barium. It was shown that these effects were not superposed upon each other, although barium had a strong catalytic action on the reaction. (author)

  1. Effects of irradiation fluence and creep on fracture toughness of 347/348 stainless steel

    International Nuclear Information System (INIS)

    Haggag, F.M.; Server, W.L.; Reuter, W.G.; Beeston, J.M.

    1984-01-01

    The postirradiation fracture toughness of Type 347/348 stainless steel was investigated using 5.08-mm thick three-point bend specimens tested at 427 0 C. The J/sub Ic/ values were determined using the single-specimen unloading compliance technique in accordance with ASTM E 813-81. Equivalent values of plane strain fracture toughness, K/sub Ic/, were computed from experimentally determined J/sub Ic/ values for several fluence levels ranging from 2.3 to 4.8 x 10 22 n/cm 2 (E > 1.0 MeV) and for irradiation creep of 0.0, 0.6, 1.1, and 1.8%. The test matrix involved four variables: fluence, creep, helium content, and heat-to-heat variation. Results show that an interpolated trend exists, i.e., K/sub Ic/ decreases with increasing combinations of fluence, creep, and helium content. These results also suggest that irradiation creep has less effect on reducing K/sub Ic/ than has been suggested previously

  2. Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

    International Nuclear Information System (INIS)

    Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

    2007-01-01

    Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

  3. Modification of graphite structure by irradiation, revealed by thermal oxidation. Examination by electronic microscopy

    International Nuclear Information System (INIS)

    Rouaud, Michel

    1969-01-01

    Based on the analysis of images obtained by electronic microscopy, this document reports the comparative study of the action of neutrons on three different graphites: a natural one (Ticonderoga) and two pyrolytic ones (Carbone-Lorraine and Raytheon). The approach is based on the modification of features of thermal oxidation of graphites by dry air after irradiation. Different corrosion features are identified. The author states that there seems to be a relationship between the number and shape of these features, and defects existing on the irradiated graphite before oxidation. For low doses, the feature aspect varies with depth at which oxidation occurs. For higher doses, the aspect remains the same [fr

  4. Carbowaste: treatment and disposal of irradiated graphite and other carbonaceous waste

    International Nuclear Information System (INIS)

    Von Lensa, W.; Rizzato, C.; Baginski, K.; Banford, A.W.; Bradbury, D.; Goodwin, J.; Grambow, B.; Grave, M.J.; Jones, A.N.; Laurent, G.; Pina, G.; Vulpius, D.

    2014-01-01

    The European Project on 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)' addressed the retrieval, characterization, treatment, reuse and disposal of irradiated graphite with the following main results: - I-graphite waste features significantly depend on the specific manufacture process, on the operational conditions in the nuclear reactor (neutron dose, atmosphere, temperature etc.) and on radiolytic oxidation leading to partial releases of activation products and precursors during operation. - The neutron activation process generates significant recoil energies breaking pre-existing chemical bonds resulting in dislocations of activation products and new chemical compounds. - Most activation products exist in different chemical forms and at different locations. - I-graphite can be partly purified by thermal and chemical treatment processes leaving more leach-resistant waste products. - Leach tests and preliminary performance analyses show that i-graphite can be safely disposed of in a wide range of disposal systems, after appropriate treatment and/or conditioning. (authors)

  5. Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite

    International Nuclear Information System (INIS)

    Bocquet, M.; Micaud, G.

    1962-01-01

    An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient ΔΘ/ Δx in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. Δx/ΔΘ, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [fr

  6. LIMITED OXIDATION OF IRRADIATED GRAPHITE WASTE TO REMOVE SURFACE CARBON-14

    Directory of Open Access Journals (Sweden)

    TARA E. SMITH

    2013-04-01

    Full Text Available Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (14C, with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the 14C, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create COx gases, i.e. “gasify” graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS and X-ray Photoelectron Spectroscopy (XPS in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl- like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a

  7. Effects of impurity trapping on irradiation-induced swelling and creep

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, L. K.; Yoo, M. H.

    1977-12-01

    A general theory of the effects of point defect trapping on radiation-induced swelling and creep deformation rates is developed. The effects on the fraction of defects recombining, and on void nucleation, void growth and creep due to the separate processes of dislocation climb-glide and dislocation climb (the so-called SIPA mechanism) are studied. Trapping of vacancies or interstitials increases total recombination and decreases the rates of deformation processes. For fixed trapping parameters, the reduction is largest for void nucleation, less for void growth and creep due to dislocation climb-glide, and least for creep due to dislocation climb. With this formation, the effects of trapping at multiple vacancy and interstitial traps and of spatial and temporal variation in trap concentrations may be determined. Alternative pictures for viewing point defect trapping in terms of effective recombination and diffusion coefficients are derived. It is shown that previous derivations of these coefficients are incorrect. A rigorous explanation is given of the well-known numerical result that interstitial trapping is significant only if the binding energy exceeds the difference between the vacancy and interstitial migration energies, while vacancy trapping is significant even at small binding energies. Corrections which become necessary at solute concentrations above about 0.1% are described. Numerical results for a wide range of material and irradiation parameters are presented.

  8. Influence of high-intensity pulsed ion beam irradiation on the creep property of 316 L stainless steel

    International Nuclear Information System (INIS)

    Wang, X.; Zhu, X.P.; Lei, M.K.; Zhang, J.S.

    2007-01-01

    High-intensity pulsed ion beam (HIPIB) treatment is a promising technology of surface modification. In this paper, the creep property of 316 L stainless steel irradiated by HIPIB at incident energy fluence per shot of 1.1-3.4 J/cm 2 with 1-10 shots at 700 deg. C have been studied. It is found that the creep property of the treated specimens after 10 shots fluctuates greatly with increasing energy density per shot. HIPIB irradiation at energy density of 1.1 J/cm 2 and 2.3 J/cm 2 prolong the creep rupture life and reduce the steady creep rate with respect to the original specimen. In contrast, HIPIB irradiation at 3.4 J/cm 2 proves to be detrimental, causing a shorter rupture life and a faster steady creep rate. Otherwise, at a fixed irradiation intensity of 2.3 J/cm 2 per shot, the number of shots has little effect on the creep property of the treated specimens. The specimens irradiated with 1, 5 and 10 shots all have better creep property compared to the control one and the difference between them is not big. The surface morphology and phase structure in the near surface region of the specimens before and after irradiation were analyzed by using scanning electron microscope (SEM) and X-ray diffraction (XRD), respectively. It is shown that the HIPIB irradiation can smooth the surface of the specimens, which can restrain the production of the surface crack. And the presence of a preferred orientation implies the treatment creates an intense compression wave and high dislocation density in the surface layer of the irradiated specimens, which hinders the dislocations movement

  9. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  10. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  11. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  12. Deformation and fracture of irradiated polygranular pile grade A reactor core graphite

    International Nuclear Information System (INIS)

    Heard, P.J.; Wootton, M.R.; Moskovic, R.; Flewitt, P.E.J.

    2011-01-01

    Highlights: → Mechanical properties of PGA nuclear graphite specimens were tested. → Load-displacement characteristics were consistent with quasi-brittle behaviour. → Micro-cracks observed in non-linear part of load-displacement curve pre-peak load. → Micro-testing showed surface tilts consistent with twinning. → Irradiated specimens failed at 20% lower load for 15% weight loss material. - Abstract: Pile grade A (PGA) graphite is used as a moderator in UK gas cooled nuclear reactors. This is a polygranular, aggregate material with quasi-brittle behaviour. When exposed to the service environment the material is subject to radiolytic oxidation that results in mass loss and an attendant increase in porosity. In the present work both unirradiated and irradiated small specimens of PGA graphite have been subjected to diametral compression. A novel trench-probe loading method is also described that allows micro-scale specimens prepared by focused ion beam milling to be fractured in a focused ion beam work station. This allows the fracture characteristics of selected regions of the graphite microstructure to be interrogated. The load-displacement and fracture characteristics of both the unirradiated and irradiated PGA graphite are compared and shown to be consistent with quasi-brittle behaviour. In addition, surface features consistent with elastically induced twins are observed associated with filler particles of the graphite. The results are discussed with respect to the quasi-brittle behaviour of this polygranular graphite.

  13. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO 2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10 25 through 1.4 × 10 25 (/m 2 , E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  14. Irradiation of graphite cloth at various temperatures with deutrons and helium ions

    International Nuclear Information System (INIS)

    Ekern, R.; Das, S.K.; Kaminsky, M.

    1975-01-01

    Graphite cloth samples were irradiated with 100 keV deuterons and 4 He + ions at room temperature and at elevated temperatures. Scanning electron microscopy was used to examine the surfaces of irradiated and unirradiated graphite fibers. Irradiation at room temperature with 4 He + to a total dose of 3.1 x 10 18 ions cm -2 produces considerable flaking of individual fibers, which is not observed on unirradiated fibers. Identical irradiations at 400 0 and 800 0 with 4 He + did not produce any detectable flaking or other surface damage. The elevated temperatures apparently prevent an accumulation of helium in localized areas which in turn could cause flaking in near surface regions. Results obtained for deuteron bombardment of graphite cloth at room temperature and at 600 0 C are also discussed

  15. Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

    International Nuclear Information System (INIS)

    Arsene, Carmen

    2009-01-01

    The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

  16. Temperature dependence of the thermal expansion of neutron-irradiated pyrolytic carbon and graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1988-01-01

    The effects of neutron irradiation and annealing on the temperature dependence of the linear thermal expansion of pyrolytic carbon and graphite were investigated after irradiation at 930-1280 0 C to a maximum neutron fluence of 2.84 x 10 25 m -2 (E > 29 fJ). After irradiation, little change in the thermal expansion of pyrolytic graphite was observed. However, as-deposited pyrolytic carbon showed an increase in thermal expansion in the perpendicular direction, a decrease in the direction parallel to the deposition plane, and also an increase in the anisotropy of the thermal expansion. Annealing at 2000 0 C did not cause any effective changes for irradiated specimens of either as-deposited pyrolytic carbon or pyrolytic graphite. (author)

  17. Creep and creep rupture of laminated graphite/epoxy composites. Ph.D. Thesis. Final Report, 1 Oct. 1979 - 30 Sep. 1980

    Science.gov (United States)

    Dillard, D. A.; Morris, D. H.; Brinson, H. F.

    1981-01-01

    An incremental numerical procedure based on lamination theory is developed to predict creep and creep rupture of general laminates. Existing unidirectional creep compliance and delayed failure data is used to develop analytical models for lamina response. The compliance model is based on a procedure proposed by Findley which incorporates the power law for creep into a nonlinear constitutive relationship. The matrix octahedral shear stress is assumed to control the stress interaction effect. A modified superposition principle is used to account for the varying stress level effect on the creep strain. The lamina failure model is based on a modification of the Tsai-Hill theory which includes the time dependent creep rupture strength. A linear cumulative damage law is used to monitor the remaining lifetime in each ply.

  18. Irradiation creep and void swelling of two LMR heat of HT9 at ∼400 degrees C and 165 dpa

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1996-01-01

    Two nominally identical heats of HT9 ferritic-martensitic steel were produced, fabricated into pressurized tubes, and then irradiated in FFTF, using identical procedures. After reaching 165 dpa at ∼400C, small differences in strains associated with both phase-related change in lattice parameter and void swelling were observed in comparing the two heats. The creep strains, while different, exhibited the same functional relationship to the swelling behavior. The derived creep coefficients, the one associated with creep in the absence of swelling and the one directly responsive to swelling, were essentially identical for the two heats. Even more significantly, the creep coefficients for this bcc ferritic-martensitic steel appear to be very similar and possibly identical to those routinely derived from creep experiments on fcc austenitic steels

  19. AGC-2 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

    2013-08-01

    This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  20. A discussion of possible mechanisms affecting fission product transport in irradiated and unirradiated nuclear grade graphite

    International Nuclear Information System (INIS)

    Firth, M.J.

    1977-09-01

    137 Cs, 85 Sr, and sup(110m)Ag adsorption experiments were conducted on three graphite powders with differing amounts of specific basal and edge surface areas. No direct proportionality was found between the specific amounts of the isotopes adsorbed and either of the surface characteristics. There appears to be some correlation with the specific basal surface area despite the fact that each isotope behaves differently. Factors that might influence the adsorption behaviour of Cs and Ag during reactor irradiation and heat treatment of nuclear grade graphites are discussed. These include the form of Cs with the graphite surface. A model based on Cs adsorption at vacancy clusters is used to analyse adsorption experiments. A possible explanation for the behaviour of Ag through the migration of graphite impurities from the bulk of the graphite to the pore surface is also discussed. (author)

  1. A systematic study of acoustic emission from nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; McEnaney, B.

    1996-01-01

    Acoustic emission (AE) monitoring has been identified as a possible method to determine internal stresses in nuclear graphites using the Kaiser effect, i.e., on stressing a graphite that has been subject to a prior stress, the onset of AE occurs at the previous peak stress. For three nuclear graphites (PGA, IM1-24 and VNEC), AE was monitored during both monotonic and cyclic loading to failure in tensile, compressive and flexural test modes. For unirradiated graphites, the Kaiser effect was not found in cyclic loading, but a Felicity effect was observed, i.e., the onset of AE occurred below the previously applied peak stress. The Felicity effect was attributed to time-dependent relaxation and recovery processes and was characterized using a new parameter, the Recovery ratio. It was shown that AE can be used to monitor creep strain and creep recovery in graphites at zero load. The AE-time responses from these experiments were fitted to equations similar to those used for creep strain-time at elevated temperatures. The number of AE counts from irradiated graphites were greater than those from unirradiated graphites, subject to similar stresses, due to increases in porosity caused by radiolytic oxidation. A Felicity effect was also observed on cyclic loading of irradiated graphites, but no evidence for a Kaiser effect was found for irradiated graphites loaded monotonically to failure. Thus internal stresses in irradiated graphites could not be measured using AE. This was attributed to relaxation and recovery processes that occur between removing the irradiated graphite from the reactor and AE testing. This work indicated that AE monitoring is not a suitable technique for measuring internal stresses in irradiated graphite. (author). 19 refs, 6 figs, 6 tabs

  2. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  3. Updating irradiated graphite disposal: Project 'GRAPA' and the international decommissioning network.

    Science.gov (United States)

    Wickham, Anthony; Steinmetz, Hans-Jürgen; O'Sullivan, Patrick; Ojovan, Michael I

    2017-05-01

    Demonstrating competence in planning and executing the disposal of radioactive wastes is a key factor in the public perception of the nuclear power industry and must be demonstrated when making the case for new nuclear build. This work addresses the particular waste stream of irradiated graphite, mostly derived from reactor moderators and amounting to more than 250,000 tonnes world-wide. Use may be made of its unique chemical and physical properties to consider possible processing and disposal options outside the normal simple classifications and repository options for mixed low or intermediate-level wastes. The IAEA has an obvious involvement in radioactive waste disposal and has established a new project 'GRAPA' - Irradiated Graphite Processing Approaches - to encourage an international debate and collaborative work aimed at optimising and facilitating the treatment of irradiated graphite. Copyright © 2017 Elsevier Ltd. All rights reserved.

  4. Role of Defects in Swelling and Creep of Irradiated SiC

    Energy Technology Data Exchange (ETDEWEB)

    Szlufarska, Izabela [Univ. of Wisconsin, Madison, WI (United States); Voyles, Paul [Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-16

    Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and

  5. Irradiation creep at temperatures of 400 degrees C and below for application to near-term fusion devices

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Gibson, L.T.; Mansur, L.K.

    1996-01-01

    To study irradiation creep at 400 degrees C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400 degrees C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330 degrees C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys

  6. An analytical study on porosity changes of nuclear graphites under high temperature irradiations

    International Nuclear Information System (INIS)

    Arai, T.

    1996-01-01

    A quantitative description of the changing pore structure, based on some radiation damage mechanisms, may introduce a physically appropriate method for lifetime assessment of graphite fuel and moderator components. Recently Brocklehurst and Kelly have analyzed well-characterized data on dimensional changes of UK reactor graphites to quantify volumetric and linear pore generation terms. The analysis (B/K theory) has demonstrated that a crystal strain parameter X T , depending on irradiation temperature and fluence, is suitable for defining structure factors, which relate changes in microstructure with those in macroscopic properties of a family of nuclear graphites. Graphite components in high temperature reactors are subjected to higher temperatures well above 1000 deg. C, which accelerate pore generation. Their mechanical integrity will suffer from the deterioration, resulting in a reduced lifetime. Previous design considerations on the dimensional change behavior have been based on an empirical approach using measured data obtained in a number of irradiation experiments. A large variety of experimental data have been utilized to develop a general phenomenological model(Graphite Damage Model, GDM) for predicting engineering properties of nuclear graphites. The present study tries to combine the B/K theory with the GDM prediction with a view to characterizing porosity changes at high temperatures of some graphites from different manufacturing routes. The dimensional change data in the literature are analyzed by the GDM to obtain their analytical presentation as a function of temperature and fluence. The results are used to derive an X T function and pore volume change as a function of X T for each grade of graphite. The resulting porosity changes are compared between different kinds of graphites. 13 refs, 6 figs, 3 tabs

  7. Mass removal by oxidation and sublimation of porous graphite during fiber laser irradiation

    Science.gov (United States)

    Phillips, Grady T.; Bauer, William A.; Fox, Charles D.; Gonzales, Ashley E.; Herr, Nicholas C.; Gosse, Ryan C.; Perram, Glen P.

    2017-01-01

    The various effects of laser heating of carbon materials are key to assessing laser weapon effectiveness. Porous graphite plates, cylinders, and cones with densities of 1.55 to 1.82 g/cm3 were irradiated by a 10-kW fiber laser at 0.075 to 3.525 kW/cm2 for 120 s to study mass removal and crater formation. Surface temperatures reached steady state values as high as 3767 K. The total decrease in sample mass ranged from 0.06 to 6.29 g, with crater volumes of 0.52 to 838 mm3, and penetration times for 12.7-mm-thick plates as short as 38 s. Minor contaminants in the graphite samples produced calcium and iron oxide to be redeposited on the graphite surface. Dramatic graphite crystalline structures are also produced at higher laser irradiances. Significantly increased porosity of the sample is observed even outside the laser-irradiated region. Total mass removed increases with deposited laser energy at a rate of 4.83 g/MJ for medium extruded graphite with an apparent threshold of 0.15 MJ. At ˜3.5 kW/cm2, the fractions of the mass removed from the cylindrical samples in the crater, surrounding trench, and outer region of decreased porosity are 38%, 47%, and 15%, respectively. Graphite is particularly resistant to damage by high power lasers. The new understanding of graphite combustion and sublimation during laser irradiation is vital to the more complex behavior of carbon composites.

  8. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  9. Measurement and computation for sag of calandria tube due to irradiation creep in PHWR

    International Nuclear Information System (INIS)

    Son, S. M.; Lee, W. R.; Lee, S. K.; Lee, J. S.; Kim, T. R.; Na, B. K.; Namgung I.

    2003-01-01

    Calandria tubes and Liquid Injection Shutdown System(LISS) tubes in a Pressurized Heavy Water Reactor(PHWR) are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  10. Swift heavy ions induced irradiation effects in monolayer graphene and highly oriented pyrolytic graphite

    International Nuclear Information System (INIS)

    Zeng, J.; Yao, H.J.; Zhang, S.X.; Zhai, P.F.; Duan, J.L.; Sun, Y.M.; Li, G.P.; Liu, J.

    2014-01-01

    Monolayer graphene and highly oriented pyrolytic graphite (HOPG) were irradiated by swift heavy ions ( 209 Bi and 112 Sn) with the fluence between 10 11 and 10 14 ions/cm 2 . Both pristine and irradiated samples were investigated by Raman spectroscopy. It was found that D and D′ peaks appear after irradiation, which indicated the ion irradiation introduced damage both in the graphene and graphite lattice. Due to the special single atomic layer structure of graphene, the irradiation fluence threshold Φ th of the D band of graphene is significantly lower ( 11 ions/cm 2 ) than that (2.5 × 10 12 ions/cm 2 ) of HOPG. The larger defect density in graphene than in HOPG indicates that the monolayer graphene is much easier to be damaged than bulk graphite by swift heavy ions. Moreover, different defect types in graphene and HOPG were detected by the different values of I D /I D′ . For the irradiation with the same electronic energy loss, the velocity effect was found in HOPG. However, in this experiment, the velocity effect was not observed in graphene samples irradiated by swift heavy ions

  11. Irradiation damage in graphite. The works of Professor B.T. Kelly

    International Nuclear Information System (INIS)

    Marsden, B.J.

    1996-01-01

    The irradiation damage produced in graphite by energetic neutrons (>100eV) has been extensively studied because of the use of graphite as a moderator in thermal nuclear reactors. In recent times, graphite has been adopted as the protective tiling of the inner wall of experimental fusion systems and property changes due to fusion neutrons have become important. The late Professor B.T. Kelly reviewed the work carried out on the irradiation behaviour of graphite since the 1940s. This work is particularly timely as the scale of research into the effects of fission neutrons has been greatly reduced and many of the active researchers have retired. In recent years, new programmes of work are being formulated for the use of graphite in both the field of high temperature reactor systems and fusion systems. It is therefore important that the knowledge gained by Professor Kelly and other workers is not lost but passed on to future generations of nuclear scientists and engineers. This paper reviews Professor Kelly's last work, it also draws on the experience gained during many long discussions with Brian during the years he worked closely with the present graphite team at AEA Technology. It is hoped to publish his work in full in the near future. (author). 13 refs, 14 figs, 3 tabs

  12. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Directory of Open Access Journals (Sweden)

    N. Simos

    2017-07-01

    Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20}  p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20}  cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite

  13. Determination of creep compliance and creep-swelling coupling coefficients for neutron-irradiated titanium-modified stainless steel at ∼400 degree C

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1991-11-01

    Irradiation creep data from FFTF-MOTA at ∼400 degrees C were analyzed for nine 20% cold-worked titanium-modified type 316 stainless steels, each of which exhibits a different duration for the transient regime of swelling. One of these steels was the fusion prime candidate alloy designated PCA. The others were various developmental breeder reactor heats. The analysis was based on the assumption that the B 0 + DS creep model applies to these steels at this temperature. This assumption was found to be valid. A creep-swelling coupling coefficient of D ∼ 0.6 x 10 -2 MPa -1 was found for all steels that had developed a significant level of swelling. This result is in excellent agreement with the results of earlier studies conducted in EBR-II using annealed AISI 304L and also 10% and 20% cold-worked AISI 316 stainless steels. There appears to be some enhancement of swelling by stress, contradicting an important assumption in the analysis and leading to an apparent but misleading nonlinearity of creep with respect to stress

  14. Annealing of neutron damage in graphite irradiated and stored at room temperature

    International Nuclear Information System (INIS)

    Gray, W.J.; Thrower, P.A.

    1979-01-01

    Annealing of neutron radiation damage in graphite at the same temperature at which it was irradiated is reported. Highly oriented pyrolytic graphite samples were irradiated to fluences in the range 0.44 to 153 x 10 15 /cm 2 at room temperature using three different neutron sources with average energies of 1.5, 5.5, and 15 MeV, respectively. Following these irradiations, the C 44 elastic constants of the samples were measured several times over periods up to two years during which time sample temperatures never exceeded 30 0 C. The C 44 constants were observed to slowly decrease toward their unirradiated values with up to 40% of the irradiation-induced changes eventually annealing out

  15. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Science.gov (United States)

    Remeikis, Vidmantas; Lagzdina, Elena; Garbaras, Andrius; Gudelis, Arūnas; Garankin, Jevgenij; Plukienė, Rita; Juodis, Laurynas; Duškesas, Grigorijus; Lingis, Danielius; Abdulajev, Vladimir; Plukis, Artūras

    2018-01-01

    14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  16. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Directory of Open Access Journals (Sweden)

    Vidmantas Remeikis

    Full Text Available 14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  17. Quality assurance for the IAEA International Database on Irradiated Nuclear Graphite Properties

    International Nuclear Information System (INIS)

    Wickham, A.J.; Humbert, D.

    2006-06-01

    Consideration has been given to the process of Quality Assurance applied to data entered into current versions of the IAEA International Database on Irradiated Nuclear Graphite Properties. Originally conceived simply as a means of collecting and preserving data on irradiation experiments and reactor operation, the data are increasingly being utilised for the preparation of safety arguments and in the design of new graphites for forthcoming generations of graphite-moderated plant. Under these circumstances, regulatory agencies require assurances that the data are of appropriate accuracy and correctly transcribed, that obvious errors in the original documentation are either highlighted or corrected, etc., before they are prepared to accept analyses built upon these data. The processes employed in the data transcription are described in this document, and proposals are made for the categorisation of data and for error reporting by Database users. (author)

  18. The effect of compressive stress on the Young's modulus of unirradiated and irradiated nuclear graphites

    International Nuclear Information System (INIS)

    Oku, T.; Usui, T.; Ero, M.; Fukuda, Y.

    1977-01-01

    The Young's moduli of unirradiated and high temperature (800 to 1000 0 C) irradiated graphites for HTGR were measured by the ultrasonic method in the direction of applied compressive stress during and after stressing. The Young's moduli of all the tested graphites decreased with increasing compressive stress both during and after stressing. In order to investigate the reason for the decrease in Young's modulus by applying compressive stress, the mercury pore diameter distributions of a part of the unirradiated and irradiated specimens were measured. The change in pore distribution is believed to be associated with structural changes produced by irradiation and compressive stressing. The residual strain, after removing the compressive stress, showed a good correlation with the decrease in Young's modulus caused by the compressive stress. The decrease in Young's modulus by applying compressive stress was considered to be due to the increase in the mobile dislocation density and the growth or formation of cracks. The results suggest, however, that the mechanism giving the larger contribution depends on the brand of graphite, and in anisotropic graphite it depends on the direction of applied stress and the irradiation conditions. (author)

  19. Irradiation creep and deformation under flux of austenitic stainless steels 304 and 316; Fluage d irradiation et deformation sous flux des aciers inoxydables austenitiques 304 et 316

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, J.; Dubuisson, P. [CEA DEN-DANS/DMN/SRMA, Saclay 91191 Gif-sur-Yvette (France); Delnondedieu, M.; Massoud, J.P. [EDF R et D, MMC, Site des Renardieres 77818 Moret sur Loing (France); Brechet, Y. [LTPCM, BP75, 38402 St Martin d Heres (France)

    2006-07-01

    The materials constituting the PWR reactors vessels internals are submitted to a neutron flux, at a temperature between 280 and 380 C, and at mechanical solicitations. On account of the C. Pokor works, the irradiation effects are now well known; the following step is the combined study of the irradiation and the mechanical solicitation. In order to understand the mechanisms which induce the microstructural changes, irradiations have been carried out in the following experimental reactors: Osiris at 330 C until 10 dpa and BOR-60 at 330 C beyond 100 dpa. Two tests types have been studied: creep tests and deformation tests under flux. Transmission electronic microscopy analyses have allowed to quantify these microstructural changes, particularly the density and the size of the Frank loops. A behaviour law, developed by J. Besson (ENSMP) and S. Leclerq (EDF), integrating irradiation creep and plasticity of the material after irradiation, allows to describe the change of stress during the reactors tests. (O.M.)

  20. Spectroscopic study of energetic helium-ion irradiation effects on nuclear graphite tiles

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Wan; Lee, K.W. [Department of Physics, Korea University, Seoul 136-713 (Korea, Republic of); Choi, D.M.; Noh, S.J.; Kim, H.S. [Department of Applied Physics, Dankook University, Yongin 448-701 (Korea, Republic of); Lee, Cheol Eui, E-mail: rscel@korea.ac.kr [Department of Physics, Korea University, Seoul 136-713 (Korea, Republic of)

    2016-02-01

    Highlights: • Energetic helium-ion irradiation on nuclear graphite tiles studied for plasma facing components. • XPS reveals recrystallization at low dose irradiation and DLC sites at higher doses. • Raman spectroscopy reveals increasing diamond-like defects and structural deformation. • Average inter-defect distance obtained as a function of irradiation dose from Raman intensities. - Abstract: Helium ion-irradiation effects on the nuclear graphite tiles were studied in order to understand the structural modifications and damages that can be produced by fusion reaction in tokamaks. The surface morphological changes due to increasing dose of the irradiation were examined by the field-effect scanning electron microscopy, and X-ray photoelectron spectroscopy elucidated the changes in the shallow surface bonding configurations caused by the energetic irradiation. Raman spectroscopy revealed the structural defects and diamond-like carbon sites that increased with increasing irradiation dose, and the average inter-defect distance was found from the Raman peak intensities as a function of the irradiation dose.

  1. Quenchable compressed graphite synthesized from neutron-irradiated highly oriented pyrolytic graphite in high pressure treatment at 1500 °C

    Science.gov (United States)

    Niwase, Keisuke; Terasawa, Mititaka; Honda, Shin-ichi; Niibe, Masahito; Hisakuni, Tomohiko; Iwata, Tadao; Higo, Yuji; Hirai, Takeshi; Shinmei, Toru; Ohfuji, Hiroaki; Irifune, Tetsuo

    2018-04-01

    The super hard material of "compressed graphite" (CG) has been reported to be formed under compression of graphite at room temperature. However, it returns to graphite under decompression. Neutron-irradiated graphite, on the other hand, is a unique material for the synthesis of a new carbon phase, as reported by the formation of an amorphous diamond by shock compression. Here, we investigate the change of structure of highly oriented pyrolytic graphite (HOPG) irradiated with neutrons to a fluence of 1.4 × 1024 n/m2 under static pressure. The neutron-irradiated HOPG sample was compressed to 15 GPa at room temperature and then the temperature was increased up to 1500 °C. X-ray diffraction, high-resolution transmission electron microscopy on the recovered sample clearly showed the formation of a significant amount of quenchable-CG with ordinary graphite. Formation of hexagonal and cubic diamonds was also confirmed. The effect of irradiation-induced defects on the synthesis of quenchable-CG under high pressure and high temperature treatment was discussed.

  2. Irradiation damage in graphite due to fast neutrons in fission and fusion systems

    International Nuclear Information System (INIS)

    2000-09-01

    Gas cooled reactors have been in operation for the production of electricity for over forty years, encompassing a total of 56 units operated in seven countries. The predominant experience has been with carbon dioxide cooled reactors (52 units), with the majority operated in the United Kingdom. In addition, four prototype helium cooled power plants were operated in the United States and Germany. The United Kingdom has no plans for further construction of carbon dioxide units, and the last helium cooled unit was shutdown in 1990. However, there has been an increasing interest in modular helium cooled reactors during the 1990s as a possible future nuclear option. Graphite is a primary material for the construction of gas cooled reactor cores, serving as a low absorption neutron moderator and providing a high temperature, high strength structure. Commercial gas cooled reactor cores (both carbon dioxide cooled and helium cooled) utilise large quantities of graphite. The structural behaviour of graphite (strength, dimensional stability, susceptibility to cracking, etc.) is a complex function of the source material, manufacturing process, chemical environment, and temperature and irradiation history. A large body of data on graphite structural performance has accumulated from operation of commercial gas cooled reactors, beginning in the 1950s and continuing to the present. The IAEA is supporting a project to collect graphite data and archive it in a retrievable form as an International Database on Irradiated Nuclear Graphite Properties, with limited general access and more detailed access by participating Member States. Because of the large size of the database, the complexity of the phenomena and the number of variables involved, a general understanding of graphite behaviour is essential to the understanding and use of the data

  3. Influence of neutron irradiation on the microstructure of nuclear graphite: An X-ray diffraction study

    Science.gov (United States)

    Zhou, Z.; Bouwman, W. G.; Schut, H.; van Staveren, T. O.; Heijna, M. C. R.; Pappas, C.

    2017-04-01

    Neutron irradiation effects on the microstructure of nuclear graphite have been investigated by X-ray diffraction on virgin and low doses (∼ 1.3 and ∼ 2.2 dpa), high temperature (750° C) irradiated samples. The diffraction patterns were interpreted using a model, which takes into account the turbostratic disorder. Besides the lattice constants, the model introduces two distinct coherent lengths in the c-axis and the basal plane, that characterise the volumes from which X-rays are scattered coherently. The methodology used in this work allows to quantify the effect of irradiation damage on the microstructure of nuclear graphite seen by X-ray diffraction. The results show that the changes of the deduced structural parameters are in agreement with previous observations from electron microscopy, but not directly related to macroscopic changes.

  4. Neutron irradiation induced microstructural changes in NBG-18 and IG-110 nuclear graphites

    Energy Technology Data Exchange (ETDEWEB)

    Karthik, Chinnathambi [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Kane, Joshua [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Idaho National Lab. (INL), Idaho Falls, ID (United States); Butt, Darryl P. [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Windes, William E. [Idaho Falls, ID (United States) Center for Advanced Energy Studies; Idaho National Lab. (INL), Idaho Falls, ID (United States); Ubic, Rick [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies

    2015-05-01

    This paper reports the neutron-irradiation-induced effects on the microstructure of NBG-18 and IG-110 nuclear graphites. The high-temperature neutron irradiation at two different irradiation conditions was carried out at the Advanced Test Reactor National User Facility at the Idaho National Laboratory. NBG-18 samples were irradiated to 1.54 dpa and 6.78 dpa at 430 °C and 678 °C respectively. IG-110 samples were irradiated to 1.91 dpa and 6.70 dpa at 451 °C and 674 °C respectively. Bright-field transmission electron microscopy imaging was used to study the changes in different microstructural components such as filler particles, microcracks, binder and quinoline-insoluble (QI) particles. Significant changes have been observed in samples irradiated to about 6.7 dpa. The closing of pre-existing microcracks was observed in both the filler and the binder phases. The binder phase exhibited substantial densification with near complete elimination of the microcracks. The QI particles embedded in the binder phase exhibited a complete microstructural transformation from rosettes to highly crystalline solid spheres. The lattice images indicate the formation of edge dislocations as well as extended line defects bridging the adjacent basal planes. The positive climb of these dislocations has been identified as the main contributor to the irradiation-induced swelling of the graphite lattice.

  5. Erosion of CFC, pyrolytic and boronated graphite under short pulsed laser irradiation

    International Nuclear Information System (INIS)

    Kraaij, G.J.; Bakker, J.; Stad, R.C.L. van der

    1992-07-01

    The effect of short pulsed laser irradiation of '0/3' ms and up to 10 MJ/m 2 on different types of carbon base materials is described. These materials are investigated as candidate protection materials for the Plasma Facing Components of NET/ITER. These materials are: carbon fibre composite graphite, pyrolytic graphite and boronated graphite. The volume of the laser induced craters was measured with an optical topographic scanner, and these data are evaluated with a simple model for the erosion. As a results, the enthalpy of ablation is estimated as 30±3 MJ/kg. A comparison is made with finite element numerical calculations, and the effect of lateral heat transfer is estimated using an analytical model. (author). 8 refs., 23 figs., 4 tabs

  6. Post-irradiation creep properties of four plates and two forgings DIN 1.4948 steel from the SNR-300 permanent primary structures

    International Nuclear Information System (INIS)

    Schaaf, B. van der.

    1987-01-01

    The safety authorities, involved in the licensing procedure of the SNR-300, have required the determination of the irradiation effect on the heat-to-heat variation of tensile and creep properties of Werkst. No. DIN 1.4948 austenitic stainless steel. These data are lacking in the present codes and they are necessary for the design and safety considerations of the permanent structures. Results are presented of about 200 tests on irradiated and unirradiated material of 6 heats used in the production of the SNR-300 permanent structures. After irradiation in the HFR-Petten to neutron fluences relevant for the SNR-300 service conditions post-irradiation tensile and creep tests (up to 10,000 hrs rupture time) were performed in the temperature range 723 K to 923 K. All heats are embrittled by irradiation resulting in reduction of rupture times, creep strength and ultimate tensile strength. The considerable reduction is attributed to helium enhanced intergranular creep crack growth, which reduces the ductility and strength, but does not affect the creep rate. The variation of tensile and creep properties is large and independent of irradiation. The minimum derived creep strength in irradiated condition drops below the values expected in the ASME Code and VdTuV Blatt. In design and safety analyses the irradiation effect on creep properties must be accounted for with an appropriate reduction factor. The predictions given, have to be verified with long-term creep tests and parts of the SNR surveillance programme. 172 figs.; 17 refs.; 58 tables

  7. Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

    2015-12-15

    Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

  8. Micro to nanostructural observations in neutron irradiated nuclear graphites PCEA and PCIB

    Science.gov (United States)

    Freeman, H. M.; Mironov, B. E.; Windes, W.; Alnairi, M. M.; Scott, A. J.; Westwood, A. V. K.; Brydson, R. M. D.

    2017-08-01

    The neutron irradiation-induced structural changes in nuclear grade graphites PCEA and PCIB were investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), Raman spectroscopy, transmission electron microscopy (TEM), selected area electron diffraction (SAED) and electron energy loss spectroscopy (EELS). The graphite samples were irradiated at the Advanced Test Reactor at the Idaho National Laboratory. Received doses ranged from 1.5 to 6.8 displacements per atom and irradiation temperatures varied between 350 °C and 670 °C. XRD and Raman measurements provided evidence for irradiation induced crystallite fragmentation, with crystallite sizes reduced by 39-55%. Analysis of TEM images was used to quantify fringe length, tortuosity, and relative misorientation of planes, and indicated that neutron irradiation induced basal plane fragmentation and curvature. EELS was used to quantify the proportion of sp2 bonding and specimen density; a slight reduction in planar-sp2 content (due to the buckling basal planes and the introduction of non-six-membered rings) agreed with the observations from TEM.

  9. Long-term creep properties, including irradiation effects, of DIN 1.4948 steel from SNR-300 primary components

    International Nuclear Information System (INIS)

    Schaaf, B. van der

    1986-01-01

    Since 1968, the ''Arbeitsgemeinschaft warmfeste Staehle in VDEH'', sponsored by KfK/PSB, performs creep tests on austenitic stainless steel DIN 1.4948. Wrought materials both in the solution-annealed condition and in weldments are investigated. The DIN 1.4948 steel was finally chosen for the permanent primary structures of the SNR-300, such as the reactor vessel, the grid plate and the primary piping. The chemical specification of DIN 1.4948 steel has been narrowed for application in the SNR-300. The long-term programme at KfK on SNR-300 heats has reached times to rupture of up to 90,000 h for parent metal. At 823 K, the operating temperature of the SNR-300, the mean creep rupture strengths of standard DIN 1.4948 and SNR-300-grade DIN 1.4948 are equal, but the scatter of data for standard DIN 1.4948 parent metal is twice as large as that for SNR-300-grade parent metal. The strength of SNR-300 weldments falls within the scatter band for parent metal. At times to rupture of more than 10 4 h, all parent metals have ductilities of between 10% and 20%, whereas the welded joints show ductilities below 1%. The neutron irradiation effects on SNR-300-grade DIN 1.4948 heats and weldments have been studied with times to rupture of up to 10,000 h in the range from 723 K to 923 K. After irradiation the creep strength is reduced to values below the minimum stress rupture values according to the ASME B + PV Code. The ductility values of parent metal after irradiation are in the range of 4% to 10%. The ductility of welded joints is sometimes below 1%. With increasing Larson-Miller parameter the reduction in creep strength decreases. Therefore, long-term post-irradiation creep tests have been started at KfK and ECN in order to verify experimentally the creep strength reduction factor due to irradiation, for times to rupture of between 10 4 and 5x10 4 h. (author)

  10. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  11. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    Science.gov (United States)

    Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

    2015-12-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

  12. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  13. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  14. Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging

    Science.gov (United States)

    Bodey, A. J.; Mileeva, Z.; Lowe, T.; Williamson-Brown, E.; Eastwood, D. S.; Simpson, C.; Titarenko, V.; Jones, A. N.; Rau, C.; Mummery, P. M.

    2017-06-01

    Nuclear graphite is used as a neutron moderator in fission power stations. To investigate the microstructural changes that occur during such use, it has been studied for the first time by X-ray microtomography with in situ heating and compression. This experiment was the first to involve simultaneous heating and mechanical loading of radioactive samples at Diamond Light Source, and represented the first study of radioactive materials at the Diamond-Manchester Imaging Branchline I13-2. Engineering methods and safety protocols were developed to ensure the safe containment of irradiated graphite as it was simultaneously compressed to 450N in a Deben 10kN Open-Frame Rig and heated to 300°C with dual focused infrared lamps. Central to safe containment was a double containment vessel which prevented escape of airborne particulates while enabling compression via a moveable ram and the transmission of infrared light to the sample. Temperature measurements were made in situ via thermocouple readout. During heating and compression, samples were simultaneously rotated and imaged with polychromatic X-rays. The resulting microtomograms are being studied via digital volume correlation to provide insights into how thermal expansion coefficients and microstructure are affected by irradiation history, load and heat. Such information will be key to improving the accuracy of graphite degradation models which inform safety margins at power stations.

  15. Path dependent models to predict property changes in graphite irradiated at changing irradiation temperatures

    CSIR Research Space (South Africa)

    Kok, S

    2010-10-01

    Full Text Available Property changes occur in materials subjected to irradiation. The bulk of experimental data and associated empirical models are for isothermal irradiation. The form that these isothermal models take is usually closed form expressions in terms...

  16. Irradiation effects on thermal shock resistance and its fracture toughness of HTGR graphites

    International Nuclear Information System (INIS)

    Sato, Sennosuke; Imamura, Yoshio; Kawamata, Kiyohiro; Awaji, Hideo; Oku, Tatsuo.

    1979-01-01

    This paper describes changes in the thermal shock resistance delta = σsub(t)k / E alpha (σsub(t): tensile strength, k: thermal conductivity, E: Young's modulus, alpha : coefficient of thermal expansion) and the thermal shock fracture toughness delta = K sub( ic)k / E alpha (K sub( ic): fracture toughness value of the mode I) in addition to usual mechanical properties including the diametral compressive strength and fracture toughness of four varieties of graphite (IM2-24, 7477, H327 and SMG) for the high temperature gas-cooled reactor due to neutron irradiations of (1.6 -- 2.3) x 10 21 n/cm 2 ( gt 0.18 MeV) at 600 -- 850 0 C. These experiments are carried out by means of our recently developed techniques using small disk type specimens which are very effective for a capsule irradiation in the JMTR. Both the thermal shock resistance and the thermal shock fracture toughness of graphites after irradiation are expressed to decrease remarkably in contrast with the increase of the usual mechanical strength. (author)

  17. IAEA International Database on Irradiated Nuclear Graphite Properties. 6th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2004-12-01

    This report summarizes the Consultant Meeting 6th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 September 2004 at Plas Tan-Y-Bwlch, Maentwrog, Gwynedd, UK. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  18. IAEA International Database on Irradiated Nuclear Graphite Properties. 7th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2005-06-01

    This report summarizes the Consultant Meeting '7th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 March 2005 at the IAEA Headquarters, Vienna, Austria. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  19. Summary report of consultants' meeting on IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2007-06-01

    The '9th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 26-27 March 2007 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  20. Effects of high temperature neutron irradiation on the physical, chemical and mechanical properties of fine-grained isotropic graphite

    International Nuclear Information System (INIS)

    Matsuo, H.; Nomura, S.; Imai, H.; Oku, T.; Eto, M.

    1987-01-01

    Effects of neutron irradiation on the dimensional change, coefficient of thermal expansion(CTE), thermal conductivity, corrosion rate, Young's modulus and strengths were studied for the candidate graphite material IG-110 of the experimental very high temperature gas-cooled reactor(VHTR) after irradiation at 585 - 1273 deg C to neutron fluences of up to about 3 x 10 25 n/m 2 (E > 29 fJ) in the JMTR and JRR-2, and to about 7 x 10 25 n/m 2 (E > 29 fJ) in the HFR. The results were compared with the irradiation behaviors of other graphites. Dimensional shrinkage was observed in the whole irradiation temperature range, showing lower value than 2 %. The shrinkage rate showed the minimum in the irradiation temperature of around 850 deg C, followed by the increase for the samples irradiated at higher temperatures. The dimensional stability of the material was clarified to be almost the same with that of H451 graphite. The CTE, thermal resistivity and Young's modulus increased in the early stage of irradiation and then only the CTE decreased while the thermal resistivity and Young's modulus levelled off with further irradiation. The neutron fluence showing the maximum CTE shifted to the lower fluence with increasing irradiation temperature. The increases of both thermal resistivity and Young's modulus were remarkable for the samples irradiated at lower temperatures. Compressive and bending strengths measured at room temperature increased after irradiation as well. The corrosion rate with water-vapor of 0.65 % in helium at high temperatures decreased owing to irradiation and the reduction was independent of irradiation temperature and neutron fluence. The activation energy for the reaction was estimated to be the same before and after irradiation. (author)

  1. Non-isothermal irradiation creep of nickel alloys Inconel 706 and PE-16

    International Nuclear Information System (INIS)

    Gilbert, E.R.; Chin, B.A.

    1984-06-01

    The results of in-reactor step temperature change experiments conducted on two nickel alloys, PE-16 and Inconel 706, were evaluated to determine the creep behavior under nonisothermal conditions. The effect of the temperature changes was found to be significantly different for the two alloys. Following a step temperature change, the creep rate of PE-16 adjusted to the rate found in isothermal tests at the new temperature. In contrast for Inconel 706, a reduction in temperature from 540 to 425 0 C produced a 300% increase in creep above that measured at 540 0 C in isothermal tests. The response of in-reactor creep in Inconel 706 to temperature changes was attributed to the dissolution of the gamma double-prime phase and subsequent loss of precipitation-strengthening at temperatures below 500 C

  2. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kohnert, Aaron A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dasgupta, Dwaipayan [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-23

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

  3. Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

    Energy Technology Data Exchange (ETDEWEB)

    Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

    2017-02-10

    The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

  4. Input Correlations for Irradiation Creep of FeCrAl and SiC Based on In-Pile Halden Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-05-01

    Swelling and creep behavior of wrought FeCrAl alloys and CVD-SiC, two candidate accident tolerant fuel cladding materials, are being examined using in-pile tests at the Halden reactor. The outcome of these tests are material property correlations that are inputs into fuel performance analysis tools. The results are discussed and compared with what is available in literature from irradiation experiments in other reactors or out-of-pile tests. Specific recommendation on what correlations should be used for swelling, thermal, and irradiation creep for each material are provided in this document.

  5. TSX graphite for extended use in the N-Reactor

    International Nuclear Information System (INIS)

    Kennedy, C.R.

    1985-08-01

    This report reviews the limited amount of irradiation data available for grade TSX graphite with the purpose of obtaining reasonable estimates of material behavior. The results are enhanced by obtaining generalized behavior characteristics demonstrated by similar grades of graphite, such as CSF, AGOT, and PGA. Intent of this work is to furnish the necessary coefficients to describe the material behavior for inclusion in the constitutive equations for the anisotropic graphite grade TSX. Estimates of the free-dimensional changes of TSX graphite as a function of temperature and fluence have been made and shown to be in good agreement with the data. The effects of irradiation on other physical properties, such as elastic moduli, conductivity, and coefficient of thermal expansion, are also described. The irradiation creep characteristics of TSX graphite are also estimated on the basis of data for similar grades of graphite in the US and Europe. Crude approximations of stresses generated in the keyed structure were made to demonstrate the magnitude of the problem. The results clearly predict that the filler-block keys will fail and the tube-block keys will not. It is also indicated that the overall stack height growth will be increased by 25 to 38 mm (1 to 1.5 in.) because of creep

  6. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  7. Tritium retention in neutron irradiated graphites, CFC's and doped C composites

    International Nuclear Information System (INIS)

    Kwast, H.; Muis, R.P.; Boshoven, J.G.

    1993-04-01

    Experiments have been performed to determine the conditions for loading graphite samples with tritium by exposure to a tritium containing gas. This is necessary to investigate the retention of tritium in irradiated graphite. The influence of exposure temperature, pressure and time as well as the sample size has been investigated. The exposure temperatures ranged from 750 C to 1150 C. The exposure time was varied from 0 to 24 hours, while total pressures of about 700 mbar to about 1000 mbar were used. So far the exposure gas was helium with 100 ppm H 2 and contained about 2.5 μCi/ml tritium at NTP, which is about 1 vppm. The effect of composition of the exposure gas and tritium concentration has still to be determined. Also longer exposure times than 24 h have to be applied to see whether or not saturation will occur at 850 C. The tritium retention of the unirradiated materials S 1611, A 05, CL 5890, Pfizer Pyrolitic and Sepcarb N112 were 18.5, 11.6, 7.8, 7.8 and 3.2 μCi/g, respectively. The exposure conditions were: 850 C, 10 hours and 800 mbar, using He/H 2 /T 2 as exposure gas. Ot the tested materials Sepcarb N112 has the lowest retention of tritium. (orig.)

  8. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunmei [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Ding, Shurong, E-mail: dsr1971@163.com [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Zhang, Xunchao; Wang, Canglong; Yang, Lei [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-12-15

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained. - Highlights: • A grain-scale gas swelling model considering the development of recrystallization and resolution is adopted for particles. • The influence of fission-gas-induced porosity is considered in the constitutive relations for particles. • A simulation method is developed for the multi-scale thermo-mechanical behavior. • The effects of fuel particle size and fission-fragment-enhanced irradiation creep are investigated in pellets.

  9. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement en fluage des alliages de zircomium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Ribis, J

    2007-12-15

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  10. Surface structure modification of single crystal graphite after slow, highly charged ion irradiation

    Science.gov (United States)

    Alzaher, I.; Akcöltekin, S.; Ban-d'Etat, B.; Manil, B.; Dey, K. R.; Been, T.; Boduch, P.; Rothard, H.; Schleberger, M.; Lebius, H.

    2018-04-01

    Single crystal graphite was irradiated by slow, highly charged ions. The modification of the surface structure was studied by means of Low-Energy Electron Diffraction. The observed damage cross section increases with the potential energy, i.e. the charge state of the incident ion, at a constant kinetic energy. The potential energy is more efficient for the damage production than the kinetic energy by more than a factor of twenty. Comparison with earlier results hints to a strong link between early electron creation and later target atom rearrangement. With increasing ion fluence, the initially large-scale single crystal is first transformed into μ m-sized crystals, before complete amorphisation takes place.

  11. Investigation of corrosion resistance of graphite under electron irradiation in the oxygen flow at the temperatures 600...800 deg C

    International Nuclear Information System (INIS)

    Zelenskij, V.F.; Odejchuk, N.P.; Ryzhov, V.P.; Borisenko, V.N.; Gamov, V.O.; Lyashchenko, A.N.; Ulybkin, A.L.; Yakovlev, V.K.

    2013-01-01

    In work results of researches of corrosion resistance of graphite samples by grades MPG, ARV and GSP (graphite bonded pyrocarbon) in oxygen flow at the temperatures of ∼ 600 and ∼ 800 deg C under the influence of electron irradiation at the accelerator ELIAS. Established that the oxidation process of graphite with the increasing temperature goes significantly more intensively and the oxidation rate increases in 6...8 times. It is shown that the best corrosion resistance under irradiation in the investigated temperature range has graphite GSP with density 1.77...1.9 g/cm 3 manufacturing of NSC KIPT

  12. Studies on mechanical properties of graphites for HTGR

    International Nuclear Information System (INIS)

    Oku, T.; Eto, M.; Fujisaki, K.; Yoda, S.; Ishiyama, S.; Sugihara, T.

    1982-01-01

    Recent research on the mechanical properties of HTGR graphites at JAERI is reviewed. The mechanical properties of graphites are required for predicting the stresses induced in the core graphite structures during reactor operation and for evaluating non-failure probabilities of the graphite structures. In this paper, effects of irradiation, stress and oxidation on the mechanical properties and fatigue properties of petroleum coke semi-isotropic and isotropic graphites for HTGRs are primarly described. Young's modulus and bend strength before and after neutron irradiation have been measured to examine irradiation effects on a fracture criterion of graphites. Two kinds of relationships are found between the bend strength and Young's modulus, depending upon the irradiation temperature. Changes in Young's modulus after irradiation are found to be different from those after irradiation creep deformation. Young's modulus under compressive stress is equivalent to that at the onset of unloading. Oxidation gives rise to the decreases in density and modulus, and also brings about a strength degradation. Tension-compression fatigue strengths are obtained and arranged successfully using statistical trivariant method with tension-tension fatigue strength data

  13. Post-irradiation tensile, creep and creep rupture data of DIN 1.4948 steel weld metal, heat-affected zones and welded joints of the SNR-300 permanent primary structures

    International Nuclear Information System (INIS)

    Schaaf, B. van der.

    1987-01-01

    DIN 1.4948 (Type 304) austenitic stainless steel has been tested for use in the SNR-300. Full weld metal specimens from a wide groove weld have been subjected to tensile and creep tests in as-deposited and stress-relieved condition. Heat-affected zones were simulated in full size standard specimens by subjecting raw cylinders to temperature cycles representative for the heat-affected zone (HAZ). The welded joints were taken from qualification welds from components, such as reactor vessel and core shield, built for the SNR-300. Half of all the specimens were irradiated in the HFR, Petten at 823 K submerged in sodium to a total neutron fluence, representative for end of life fluences of SNR-300 reactor vessel and shield tank. The tensile and creep properties of as-deposited and stress-relieved weld metal are not affected by the neutron irradiation. The stress relief treatment of 3 h at 1175 K improves both tensile and creep ductility, without affecting strength too much, by reducing the dislocation density in weld metal and HAZ and coarsening the carbides in weld metal. The HAZ shows a pronounced decrease in post-irradiation tensile ductility, strength, creep strength and ductility. Ductility of weld metal can be improved by stress-relieving and adapting the chemical composition of the filler metal. In order to prevent the HAZ to become the weakest link of a welded joint after irradiation grain growth and boron precipitation on grain boundaries must be suppressed. Low heat input welding techniques such as electron beam welding and automated gas metal arc welding might offer a solution for HAZ post-irradiation embrittlement. 144 figs.; 20 refs.; 45 tables

  14. Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2018-04-01

    Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

  15. NGNP Graphite Selection and Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, T.; Bratton, R.; Windes, W.

    2007-09-30

    The nuclear graphite (H-451) previously used in the United States for High-Temperature Reactors (HTRs) is no longer available. New graphites have been developed and are considered suitable candidates for the Next-Generation Nuclear Plant (NGNP). A complete properties database for these new, available, candidate grades of graphite must be developed to support the design and licensing of NGNP core components. Data are required for the physical, mechanical (including radiation-induced creep), and oxidation properties of graphites. Moreover, the data must be statistically sound and take account of in-billet, between billets, and lot-to-lot variations of properties. These data are needed to support the ongoing development1 of the risk-derived American Society of Mechanical Engineers (ASME) graphite design code (a consensus code being prepared under the jurisdiction of the ASME by gas-cooled reactor and NGNP stakeholders including the vendors). The earlier Fort St. Vrain design of High-Temperature Reactor (HTRs) used deterministic performance models for H-451, while the NGNP will use new graphite grades and risk-derived (probabilistic) performance models and design codes, such as that being developed by the ASME. A radiation effects database must be developed for the currently available graphite materials, and this requires a substantial graphite irradiation program. The graphite Technology Development Plan (TDP)2 describes the data needed and the experiments planned to acquire these data in a timely fashion to support NGNP design, construction, and licensing. The strategy for the selection of appropriate grades of graphite for the NGNP is discussed here. The final selection of graphite grades depends upon the chosen reactor type and vendor because the reactor type (pebble bed or prismatic block) has a major influence on the graphite chosen by the designer. However, the time required to obtain the needed irradiation data for the selected NGNP graphite is sufficiently

  16. The effects of neutron irradiation on various fuels used in gas graphite reactor systems

    International Nuclear Information System (INIS)

    Englander, M.

    1964-01-01

    The behaviour of natural uranium based fuels in form of rods and tubes has been examined after irradiation in several French reactors. Two categories of uranium alloys have thus been studied: on one hand alloys containing from 0,5 to 3 p.100 by weight of molybdenum, and on the other hand alloys obtained by addition to nuclear purity uranium, of less than 1500 ppm of metallic elements such as iron, aluminium, chromium and silicon. The tests have been carried out with two types of elements: one canned with a magnesium alloy and cooled by carbon dioxide under pressure (in G 2 G 3), the other canned with aluminium and cooled by heavy water (in EL 3). Each of these elements had the same outside form as the usual elements in these reactors, but the enrichment and thickness of the fuel were adapted in order to meet different irradiation conditions. The post-irradiation examinations described, thereafter involved techniques of metrology, of radiography, of density measurements before and after isochron annealings, as well as macro-graphies or optical and electronic micrographies. The experimental results thus obtained reflect the changes in the internal and external structures of the full-size fuel elements; they are discussed in terms of the combined effects of pressure, temperature and burn-up on the morphological state, on the isotropy and on the mechanical characteristics of the initial polycrystalline aggregate. An analysis of the results makes it possible to deduce the various advantages of un-enriched metallic fuels designed to operate with a hot-point of around 600 C in reactors of the gas-graphite type, and for a length of time corresponding to an output of 4500 MW day/metre ton. (author) [fr

  17. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  18. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    Science.gov (United States)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for "Uranium Naturel-Graphite-Gaz") to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the total amount produced

  19. Effect of neutron irradiation on creep, fatigue and tensile properties of stainless steel type DIN 1.4948 (similar to AISI 304)

    International Nuclear Information System (INIS)

    Elen, J.D.; Vries, M.I. de; Schaff, B. van der; Staal, H.U.

    1978-03-01

    As a contribution to the German-Belgian-Dutch fast breeder project SNR-300 a mechanical testing programme is being performed at ECN to determine the effects of neutron irradiation on the mechanical properties of the DIN 1.4948 construction steel of the SNR-300 reactor vessel and internal components. Irradiations of plate and weld samples were performed at 723 K and 823 K to thermal neutron fluences of 6 x 10 18 n.cm -2 and 2 x 10 20 n.cm -2 in core positions of the High Flux Reactor at Petten at thermal to fast flux density ratios of about 0.6. Postirradiation testing comprises tensile testing at strain rates from 6 x 10 -6 s -1 to 6 s -1 , creep measurements up to 10.000 h rupture time and low cycle fatigue at strain ranges from 0.6% to 2% and a strain rate of 3 x 10 -3 s -1 . The major effect observed is high temperature embrittlement due to helium produced by the 10 B (n,α) 7 Li reaction in the 14 ppm boron containing steel used for the experiments. The creep rupture time of plate material at 823 K is reduced to 10% of its original value by irradiation to the lower fluence and the creep strength is decreased by 60 MN.m -2 . The total creep strain of weld samples is reduced to values of 0.3% to 1.5%

  20. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program)

    International Nuclear Information System (INIS)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability

  1. Radiation damage in graphite

    CERN Document Server

    Simmons, John Harry Walrond

    1965-01-01

    Nuclear Energy, Volume 102: Radiation Damage in Graphite provides a general account of the effects of irradiation on graphite. This book presents valuable work on the structure of the defects produced in graphite crystals by irradiation. Organized into eight chapters, this volume begins with an overview of the description of the methods of manufacturing graphite and of its physical properties. This text then presents details of the method of setting up a scale of irradiation dose. Other chapters consider the effect of irradiation at a given temperature on a physical property of graphite. This

  2. AGC 2 Irradiated Material Properties Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  3. AGC-2 Specimen Post Irradiation Data Package Report

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

  4. Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Pinhero, Patrick [Univ. of Missouri, Columbia, MO (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-10

    The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be

  5. Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

    International Nuclear Information System (INIS)

    Singh, B.N.; Johansen, B.S.; Li, M.; Stubbins, J.F.

    2005-08-01

    The creep-fatigue interaction behaviour of a precipitation hardened CuCrZr alloy was investigated at 295 and 573 K. To determine the effect of irradiation a number of fatigue specimens were irradiated at 333 and 573 K to a dose level in the range of 0.2 - 0.3 dpa and were tested at room temperature and 573 K, respectively. The creep-fatigue deformation behaviour of OFHC-copper was also investigated but only in the unirradiated condition and at room temperature. The creep-fatigue interaction was simulated by applying a certain holdtime on both tension and compression sides of the cyclic loading with a frequency of 0.5 Hz. Holdtimes of up to 1000 seconds were used. Creep-fatigue experiments were carried out using strain, load and extension controlled modes of cyclic loading. In addition, a number of 'interrupted' creep-fatigue tests were performed on the prime aged CuCuZr specimens in the strain controlled mode with a strain amplitude of 0.5% and a holdtime of 10 seconds. The lifetimes in terms of the number of cycles to failure were determined at different strain and load amplitudes at each holdtime. Post-deformation microstructures was investigated using a transmission electron microscopy. The main results of these investigations are presented and their implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the case of OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction in the yield strength due to overaging heat treatments causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K

  6. Studies of mechanical properties and irradiation damage nucleation of HTGR graphites. Progress report, February 1, 1978--March 31, 1979

    International Nuclear Information System (INIS)

    Thrower, P.A.

    1979-04-01

    Irradiation damage effects to highly oriented pyrolytic graphite by 1.5, 5.5 and 15 MeV neutrons are briefly described. Studies of the effects of oxidation on the structure and strength of Stackpole 2020, Great Lakes H440 and POCO AXF-5Q graphites have been continued using H 2 O vapor as the oxidant. The effect of burn-off on compressive strength is similar to that of CO 2 , but structural effects are quite different. The kinetics of the C--CO 2 reaction are being studied using low CO 2 partial pressures in a total pressure of one atmosphere. The necessary modifications to existing apparatus have been made and experimental work is commencing

  7. Graphite Technology Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; T. Burchell; M.Carroll

    2010-10-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled High Temperature Gas Reactor (HTGR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Graphite has been used effectively as a structural and moderator material in both research and commercial high-temperature gas-cooled reactors. This development has resulted in graphite being established as a viable structural material for HTGRs. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermomechanical design of the structural graphite in NGNP is based. This Technology Development Plan outlines the research and development (R&D) activities and associated rationale necessary to qualify nuclear grade graphite for use within the NGNP reactor.

  8. IAEA International Database on Irradiated Nuclear Graphite Properties. Summary report of consultants' meeting. 12. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Chung, H.K.; Wickham, A.J.

    2010-02-01

    The 12th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 12-13 November 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  9. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1976. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-24

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs.

  10. Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors; Effets de la temperature et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucleaire: consequences pour le stockage des graphites irradies des reacteurs UNGG

    Energy Technology Data Exchange (ETDEWEB)

    Vaudey, C.E.

    2010-10-15

    This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides ({sup 14}C and {sup 36}Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of {sup 36}Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% {sup 36}Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)

  11. Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

    International Nuclear Information System (INIS)

    Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

    2017-01-01

    There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the

  12. Precision rectifier detectors for ac resistance bridge measurements with application to temperature control systems for irradiation creep experiments

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, M. G.

    1977-05-01

    The suitability of several temperature measurement schemes for an irradiation creep experiment is examined. It is found that the specimen resistance can be used to measure and control the sample temperature if compensated for resistance drift due to radiation and annealing effects. A modified Kelvin bridge is presented that allows compensation for resistance drift by periodically checking the sample resistance at a controlled ambient temperature. A new phase-insensitive method for detecting the bridge error signals is presented. The phase-insensitive detector is formed by averaging the magnitude of two bridge voltages. Although this method is substantially less sensitive to stray reactances in the bridge than conventional phase-sensitive detectors, it is sensitive to gain stability and linearity of the rectifier circuits. Accuracy limitations of rectifier circuits are examined both theoretically and experimentally in great detail. Both hand analyses and computer simulations of rectifier errors are presented. Finally, the design of a temperature control system based on sample resistance measurement is presented. The prototype is shown to control a 316 stainless steel sample to within a 0.15/sup 0/C short term (10 sec) and a 0.03/sup 0/C long term (10 min) standard deviation at temperatures between 150 and 700/sup 0/C. The phase-insensitive detector typically contributes less than 10 ppM peak resistance measurement error (0.04/sup 0/C at 700/sup 0/C for 316 stainless steel or 0.005/sup 0/C at 150/sup 0/C for zirconium).

  13. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

  14. Studies of mechanical properties and irradiation damage nucleation of HTGR graphites. Final report

    International Nuclear Information System (INIS)

    Thrower, P.A.

    1981-05-01

    Since the submission of the last report (COO-2712-6) work has concentrated on the examination of the effects of oxidation on the compressive strengths of graphites doped with iron, vanadium and calcium. The purpose of the investigation was to determine the relative effects of the impurities on the rates of oxidation in air, CO 2 and H 2 O and the resultant reduction in compressive strength

  15. Direct synthesis of graphitic mesoporous carbon from green phenolic resins exposed to subsequent UV and IR laser irradiations

    Science.gov (United States)

    Sopronyi, Mihai; Sima, Felix; Vaulot, Cyril; Delmotte, Luc; Bahouka, Armel; Matei Ghimbeu, Camelia

    2016-12-01

    The design of mesoporous carbon materials with controlled textural and structural features by rapid, cost-effective and eco-friendly means is highly demanded for many fields of applications. We report herein on the fast and tailored synthesis of mesoporous carbon by UV and IR laser assisted irradiations of a solution consisting of green phenolic resins and surfactant agent. By tailoring the UV laser parameters such as energy, pulse repetition rate or exposure time carbon materials with different pore size, architecture and wall thickness were obtained. By increasing irradiation dose, the mesopore size diminishes in the favor of wall thickness while the morphology shifts from worm-like to an ordered hexagonal one. This was related to the intensification of phenolic resin cross-linking which induces the reduction of H-bonding with the template as highlighted by 13C and 1H NMR. In addition, mesoporous carbon with graphitic structure was obtained by IR laser irradiation at room temperature and in very short time periods compared to the classical long thermal treatment at very high temperatures. Therefore, the carbon texture and structure can be tuned only by playing with laser parameters, without extra chemicals, as usually required.

  16. Swelling and in-pile creep of neutron irradiated 15Cr15NiTi austenitic steels in the temperature range of 400 to 600 deg. C

    International Nuclear Information System (INIS)

    Huebner, R.; Ehrlich, K.

    1998-01-01

    A pressurized tube experiment was carried out in the Prototype Fast Reactor (PFR) ad Dounreay in order to determine swelling, stress-induced swelling and in-pile creep of different austenitic steels. The tubes were made out of different heats of the commercial German austenitic steel DIN 1.4970 and a number of model plain Fe-15Cr-15Ni stainless steels. Special attention was paid on the influence of minor alloying elements like Si, Ti, degree of Ti/C relation and others. The maximum doses achieved are 106 dpa NRT at 420 deg. C, 81 dpa NRT 500 deg. C and 61 dpa NRT at 600 deg. C. The hoop stresses of the pressurized tubes were 0, 60 and 120 MPa at all irradiation temperatures. The length and diameter changes of the pressurized capsules have been determined at up to four intermediate stages and after irradiation. Post irradiation examinations by immersion density measurements and transmission electron microscopy (TEM) are partially done. All alloys exhibited the highest swelling values at 420 deg. C and nearly no swelling at 600 deg.C. The measurements show the large effect of the minor alloying elements upon swelling and in-pile creep. The maximum swelling suppression is achieved for DIN 1.4970 through a high Si-content and an under stoichiometric Ti/C relation (under stabilization). This yields linear swelling of 1.9% after 106 dpa NRT at 420 deg. C. The formerly observed inter correlation between swelling and in-pile creep is confirmed up to 106 dpa NRT . It can be described by an equation consisting of a SIPA term (stress induced preferential absorption) and an inter correlation term similar to the I-creep proposed by Gittus. The estimates of the stress-induced swelling using the Soderberg theorem and the length measurements are compared with the immersion density measurements and results by TEM. The immersion density measurements agree rather good with length measurements. The stress-induced linear swelling can reach values of 0.8% at 100 dpa NRT and 120 MPa hoop

  17. Visible-Light-Irradiated Graphitic Carbon Nitride Photocatalyzed Diels-Alder Reactions with Dioxygen as Sustainable Mediator for Photoinduced Electrons.

    Science.gov (United States)

    Zhao, Yubao; Antonietti, Markus

    2017-08-01

    Photocatalytic Diels-Alder (D-A) reactions with electron rich olefins are realized by graphitic carbon nitride (g-C 3 N 4 ) under visible-light irradiation and aerobic conditions. This heterogeneous photoredox reaction system is highly efficient, and the apparent quantum yield reaches a remarkable value of 47 % for the model reaction. Dioxygen plays a critical role as electron mediator, which is distinct from the previous reports in the homogeneous Ru II complex photoredox system. Moreover, the reaction intermediate vinylcyclobutane is captured and monitored during the reaction, serving as a direct evidence for the proposed reaction mechanism. The cycloaddition process is thereby determined to be the combination of direct [4+2] cycloaddition and [2+2] cycloaddition followed by photocatalytic rearrangement of the vinylcyclobutane intermediate. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. Principles and practice of a bellows-loaded compact irradiation vehicle

    Science.gov (United States)

    Byun, Thak Sang; Li, Meimei; Snead, Lance L.; Katoh, Yutai; Burchell, Timothy D.; McDuffee, Joel L.

    2013-08-01

    -furnace and in-reactor tests. The post-irradiation in-furnace measurement of load-displacement response from bellows is the main step to evaluate the stress applied to the specimen during in-reactor creep testing. The irradiation creep tests for cylindrical graphite specimens have been successfully carried out using the newly developed technology, and the results are reported in a separate publication. Some post-irradiation load-displacement measurements indicated that the radial expansion of stainless steel bellows by irradiation creep to high doses can cause significant friction between the bellows wall and the load frame. Changes in capsule design and material selection are considered to solve this issue for high dose experiments. Expect for the irradiation creep of bellows at high doses, no other major problem has been raised.

  19. WC/Co composite surface structure and nano graphite precipitate induced by high current pulsed electron beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hao, S.Z., E-mail: ebeam@dlut.edu.cn [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); Zhang, Y. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); School of Materials Science and Engineering, Shenyang University of Technology, Shenyang 110870 (China); Xu, Y. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); College of Materials Science and Engineering, Liaoning Technical University, Fuxin 123000 (China); Gey, N.; Grosdidier, T. [Université de Lorraine, Laboratoire d’Etude des Microstructures et de Mécanique des Matériaux (LEM3), CNRS UMR 7239, Ile du Saulcy, 57045 Metz (France); Université de Lorraine, Laboratoire d’Excellence on Design of Alloy Metals for Low-Mass Structure (DAMAS), Ile du Saulcy, 57045 Metz (France); Dong, C. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); Université de Lorraine, Laboratoire d’Excellence on Design of Alloy Metals for Low-Mass Structure (DAMAS), Ile du Saulcy, 57045 Metz (France)

    2013-11-15

    High current pulsed electron beam (HCPEB) irradiation was conducted on a WC-6% Co hard alloy with accelerating voltage of 27 kV and pulse duration of 2.5 μs. The surface phase structure was examined by using glancing-angle X-ray diffraction (GAXRD), scanning electron microscope (SEM) and high resolution transmission electron microscope (HRTEM) methods. The surface tribological properties were measured. It was found that after 20 pulses of HCPEB irradiation, the surface structure of WC/Co hard alloy was modified dramatically and composed of a mixture of nano-grained WC{sub 1−x}, Co{sub 3}W{sub 9}C{sub 4}, Co{sub 3}W{sub 3}C phases and graphite precipitate domains ∼50 nm. The friction coefficient of modified surface decreased to ∼0.38 from 0.6 of the initial state, and the wear rate reduced from 8.4 × 10{sup −5} mm{sup 3}/min to 6.3 × 10{sup −6} mm{sup 3}/min, showing a significant self-lubricating effect.

  20. AGC-1 Post Irradiation Examination Status

    Energy Technology Data Exchange (ETDEWEB)

    David Swank

    2011-09-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

  1. Toxicological characterization of chemicals produced from laser irradiation of graphite composite materials

    International Nuclear Information System (INIS)

    Kwan, J.

    1990-11-01

    One of the major potential hazards associated with laser machining of graphite composite materials is the toxic fumes and gases that are generated. When exposed to the intense energy of the laser beam, the organic polymer matrix of the composite material may decompose into various toxic by-products. To advance the understanding of the laser machining process from a health and safety viewpoint, this particular study consisted of the following steps: collect and analyze gaseous by-products generated during laser machining; collect particulates generated during laser machining and chemically extract them to determine the chemical species that may have absorbed or recondensed onto these particles; and review and evaluate the toxicity of the identified chemical species

  2. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  3. Nuclear graphite ageing and turnaround

    International Nuclear Information System (INIS)

    Marsden, B.J.; Hall, G.N.; Smart, J.

    2001-01-01

    Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

  4. Fabrication of SnO₂-reduced graphite oxide monolayer-ordered porous film gas sensor with tunable sensitivity through ultra-violet light irradiation.

    Science.gov (United States)

    Xu, Shipu; Sun, Fengqiang; Yang, Shumin; Pan, Zizhao; Long, Jinfeng; Gu, Fenglong

    2015-03-11

    A new graphene-based composite structure, monolayer-ordered macroporous film composed of a layer of orderly arranged macropores, was reported. As an example, SnO2-reduced graphite oxide monolayer-ordered macroporous film was fabricated on a ceramic tube substrate under the irradiation of ultra-violet light (UV), by taking the latex microsphere two-dimensional colloid crystal as a template. Graphite oxide sheets dispersed in SnSO4 aqueous solution exhibited excellent affinity with template microspheres and were in situ incorporated into the pore walls during UV-induced growth of SnO2. The growing and the as-formed SnO2, just like other photocatalytic semiconductor, could be excited to produce electrons and holes under UV irradiation. Electrons reduced GO and holes adsorbed corresponding negative ions, which changed the properties of the composite film. This film was directly used as gas-sensor and was able to display high sensitivity in detecting ethanol gas. More interestingly, on the basis of SnO2-induced photochemical behaviours, this sensor demonstrated tunable sensitivity when UV irradiation time was controlled during the fabrication process and post in water, respectively. This study provides efficient ways of conducting the in situ fabrication of a semiconductor-reduced graphite oxide film device with uniform surface structure and controllable properties.

  5. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

  6. A 3-D inelastic analysis of HTR graphite structures and a comparison with A 2-D approach

    International Nuclear Information System (INIS)

    Willaschek, J.

    1979-01-01

    In High Temperature Reactor Cores (HTR) a large number of elements are constructed of nuclear graphite. The dimensions of the graphite components are limited by stresses and strains resulting from thermal loads, irradiation induced dimensional changes and stress-dependent irradiation creep. Therefore it is necessary to examine the feasibility of design concepts with regard to the structural integrity of the material. This paper presents an analysis of a radial reflector concept for use in a 3000 MWth HTR for process heat production. This concept of a pebble bed reactor (OTTO cycle) requires reflector dimensions and shapes which have previously not been used and which may exceed acceptable stress limits. Graphite reflector elements in a HTR are subject to a high fluence of fast neutrons. The fluence varies spatially within an element. Irradiation-induced strains occur which in turn vary non-linearly with the fluence. At low fluences the graphite shrinks. With increasing fluence shrinkage is saturated and after a 'turn-around' point the graphite begins to swell. The net effect of fluence gradient and irradiation-induced strain is a 'necking' of the element which moves radially outwards with time. In this paper a three-dimensional inelastic analysis of a graphite block with the above deformation history is described. The influence of irradiation on dimensional stability and other material properties was taken into account. Numerical results were obtained with the finite-element computer code ADINA, modified at INTERATOM for the task in hand. The radial reflector block was modelled using 21-node three-dimensional continuum elements of elastic-creep material. The element stiffness matrices were calculated using the standard 2x2x2 Gauss integration; material nonlinearities with quadratic displacement functions and linearised initial strains were employed. (orig.)

  7. Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming

    International Nuclear Information System (INIS)

    Silbermann, Gwennaelle

    2013-01-01

    The dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14 C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14 C and its precursor, 14 N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14 C and 14 N by the respective ion implantation of 13 C and 14 N or 15 N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500 C) and without radiolytic corrosion, 13 C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500 C) put in contact with a gas representative of the radiolized coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13 C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13 C more weakly bound than 12 C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (≤500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13 C, implanted nitrogen migrates at 500 C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature

  8. Creep behavior of Zr-Nb alloys

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Chan; Kim, Young Suk; Cheong, Yong Mu; Kwon, Sang Chul; Kim, Sung Soo; Choo, Ki Nam [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    The creep characteristics of Zirconium alloy is affected by several parameters. Out-reactor creep increases both with an increasing amount of Nb, Sn and S contained in alpha-Zr and decreases with the increasing volume of alpha-Zr. Especially, the creep of Zr-2.5Nb alloy depends on the solubility of Nb in alpha-Zr, which is associated with the decomposition of beta-Zr. Since Zr of the hcp structure is strongly anisotropic, it shows the characteristics of texture and results in the anisotropy of creep. Due to the circumferential texture of Zr-2.5%Nb alloy (CANDU Pressure tube), the longitudinal slip is easier than the circumferential one, resulting in the high creep rate. The irradiation creep also increases with increasing neutron fluence. The neutron irradiation increases the strength of the zirconium alloys but decreases their creep strength. In contrast to the out-reactor creep, the irradiation creep is little sensitive to temperature, resulting in the lower activation energy. The most important factor to affect the in-reactor and out-reactor creep of niobium containing alloys seems to be the solution hardening by Nb or Sn which is soluble in alpha-zirconium and the texture as well. Irradiation growth is the mechanism which is caused only by the irradiation. It becomes saturated at lower fluence than the critical fluence but beyond it, shows the break-away growth. The onset of accelerated irradiation growth corresponds with the c-dislocation loop formation, though its mechanism needs better understanding. Generally, the irradiation growth of Zr-Nb alloys increases with an increase in fluence, cold working, dislocation, density and temperature, and with a decrease in the grain size. 141 refs., 59 figs., 10 tabs. (Author)

  9. Comparison of 3 MeV C + ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    Science.gov (United States)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-10-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ˜19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  10. Comparison of 3 MeV C+ ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-01-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ∼19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  11. Organic matters removal from landfill leachate by immobilized Phanerochaete chrysosporium loaded with graphitic carbon nitride under visible light irradiation.

    Science.gov (United States)

    Hu, Liang; Liu, Yutang; Zeng, Guangming; Chen, Guiqiu; Wan, Jia; Zeng, Yunxiong; Wang, Longlu; Wu, Haipeng; Xu, Piao; Zhang, Chen; Cheng, Min; Hu, Tianjue

    2017-10-01

    This study investigated the technical applicability of a combination of Phanerochaete chrysosporium (P. chrysosporium) with photocatalyst graphitic carbon nitride (g-C 3 N 4 ) for organic matters removal from landfill leachate under visible light irradiation. Photocatalyst g-C 3 N 4 was well immobilized on the hyphae surface of P. chrysosporium by calcium alginate. The typical absorption edge in visible light region for g-C 3 N 4 was at about 460 nm, and the optical absorption bandgap of g-C 3 N 4 was estimated to be 2.70 eV, demonstrating the great photoresponsive ability of g-C 3 N 4 . An optimized g-C 3 N 4 content of 0.10 g in immobilized P. chrysosporium and an optimized immobilized P. chrysosporium dosage of 1.0 g were suitable for organic matters removal. The removal efficiency of total organic carbon (TOC) reached 74.99% in 72 h with the initial TOC concentration of 100 mg L -1 . In addition, the gas chromatography coupled with mass spectrometry (GC-MS) measurements showed that immobilized P. chrysosporium presented an outstanding removal performance for almost all organic compounds in landfill leachate, especially for the volatile fatty acids and long-chain hydrocarbons. The overall results indicate that the combination P. chrysosporium with photocatalyst g-C 3 N 4 for organic matters removal from landfill leachate may provide a more comprehensive potential for the landfill leachate treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW); Recozimento do moderador de grafite do reator experimental de irradiacao (0.5MW)

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves [Centro Tecnologico do Exercito, Rio de Janeiro, RJ (Brazil). Instituto de Projetos Especiais

    1995-12-31

    This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs.

  13. Summary report of consultants meeting on IAEA International Database on Irradiated Nuclear Graphite Properties. 11. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2009-05-01

    The 11th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 25-26 March 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  14. Evaluation of plasma disruption simulating short pulse laser irradiation experiments on boronated graphites and CFCs [carbon fibre composites

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der; Klippel, H.T.; Kraaij, G.J.

    1992-12-01

    New experimental and numerical results from disruption heat flux simulations in the millisecond range with laser beams are discussed. For a number of graphites, boronated graphites and carbon fibre composites, the effective enthalpy of ablation is determined as 30 ± 3 MJ/kg, using laser pulses of about -.3 ms. The numerical results predict the experimental results rather well. No effect of boron doping on the ablation enthalpy is found. (author). 9 refs., 4 figs., 1 tab

  15. Application of the concepts of thermoactivation analysis to interpretation of the effect of irradiation on creep and long-term strength

    International Nuclear Information System (INIS)

    Geminov, V.N.; Votinov, S.N.

    1989-01-01

    Thermoactivation parameters occupy a prominent position among the characteristics of the mechanisms of deformation and damage in a metal. These include the true activation energy U 0 (the height of the activation barrier overcome by atoms during their irreversible movement) and the activation volume γ (a structure-sensitive factor). Knowledge of these parameters and of their patterns of change should assist in identifying the mechanisms and kinetics of creep strain for different conditions, since distinctive thermoactivation parameters correspond to each mechanism. The authors investigate these parameters. The paper concludes that: the use of the concept of thermoactivation parameters in conjunction with mechanism maps opens up new possibilities for analysis and identification of the mechanisms of action of irradiation or bombardment on a material; the use of this concept makes it possible to increase sharply the information yield of experimental data arrays, on the one hand, and, on the other hand, to reduce significantly the number of experiments required, through the construction of a single diagram. This is especially promising in particular for reducing the number of in-reactor tests

  16. Graphites for nuclear applications; Les graphites pour les applications nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Bonal, J.P.; Gosmain, L. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DMN), Lab. de Microscopie et d' Etudes de l' Endommagement, 91 - Gif-sur-Yvette (France)

    2006-03-15

    Being an excellent neutron moderator, graphite is used as a structural material in many nuclear reactor types. By the end of the 50's, the French gas-cooled reactor development needed manufacturing of a nuclear-grade graphite. Graphite irradiation can lead to in-lattice energy accumulation, dimensional changes and physical properties modification. Moreover, the radiolytic corrosion induced by the coolant (CO{sub 2}) may generate mechanical properties degradation. Today, French gas-cooled reactors are all in their decommissioning phase that requires the knowledge of the radiological inventory of the irradiated graphites. At present time, graphite is still foreseen as a future material for hydrogen production by high temperature gas cooled nuclear plants. In the future, graphite will be the necessary moderator material for high temperature reactors with thermal neutron spectrum dedicated to hydrogen and electricity production. (authors)

  17. Radiation and thermal effects on the time-dependent response of T300/934 graphite/epoxy

    Science.gov (United States)

    Yancey, Robert N.; Pindera, Marek-Jerzy; Slemp, Wayne; Funk, Joan G.

    1989-01-01

    Experimental studies have suggested that radiation, in conjunction with elevated temperatures, may lead to an exacerbation of the time-dependent response of such materials as T300/934 graphite/epoxy. Attention is given to the results of an investigation into such radiation/temperature effects with this composite system; creep tests were conducted on specimens of both irradiated and nonirradiated T300/934 composite and neat 934 resin specimens, at room temperature and 121 C. The radiation was of 1 MeV electrons, for a total dose of 10,000 Mrads; this simulates a 30-year exposure to radiation in geosynchronous orbit. Irradiation at elevated temperature led to a significant creep-response increase in the initial stages of loading.

  18. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  19. Graphite surveillance in N Reactor

    International Nuclear Information System (INIS)

    Woodruff, E.M.

    1991-09-01

    Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

  20. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  1. Physical mechanisms of creep

    International Nuclear Information System (INIS)

    Rojtburd, A.L.

    1976-01-01

    The review is devoted to microscopic theories of creep in the range of temperatures where it is determined by self-diffusion. Interactions of vacancies with dislocations resulting in creeping over the latter are considered in detail. It is demonstrated that even at high temperatures the creep is determined by the process which includes creeping over and sliding of dislocations. Specific features of the description of polycrystal creep are considered

  2. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Arthur, B.; Lui, Y.Y.

    1985-01-01

    The objective of this effort was to obtain data on the performance of lithium ceramic materials during fast neutron irradiation in support of solid breeder blanket designs. Li 2 O has been observed to swell (greater than or equal to 4%) under fast reactor irradiation. Fortunately, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory creep experiments were conducted between 500 to 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  3. Study by electronic microscopy of corrosion features of graphite after hot oxidation (air, 620 C)

    International Nuclear Information System (INIS)

    Jodon de Villeroche, Suzanne

    1968-01-01

    The author reports the study of corrosion features of graphite after hot oxidation in the air at 620 C. It is based on observations made by electronic microscopy. This study comes after another one dedicated to oxidation features obtained by hot corrosion of natural graphite, and aims at comparing pyrolytic graphite before and after irradiation in an atomic pile, and at performing tests on a graphite processed with ozone. After a recall of generalities about natural graphite and of some issues related to hot corrosion of natural graphite, the author presents some characteristics and features of irradiated and non-irradiated pyrolytic graphite. He reports the study of the oxidation of samples of pyrolytic graphite: production of thin lamellae, production of glaze-carbon replicates, oxidation of irradiated and of non-irradiated graphite, healing of irradiation defects, and oxidation of ozone-processed natural graphite [fr

  4. Creep of crystals

    International Nuclear Information System (INIS)

    Poirier, J.-P.

    1988-01-01

    Creep mechanisms for metals, ceramics and rocks, effect of pressure and temperature on deformation processes are considered. The role of crystal defects is analysed, different models of creep are described. Deformation mechanisms maps for different materials are presented

  5. Simultaneous consolidation and creep

    DEFF Research Database (Denmark)

    Krogsbøll, Anette

    1997-01-01

    Materials that exhibit creep under constant effective stress typically also show rate dependent behavior. The creep deformations and the rate sensitive behavior is very important when engineering and geological problems with large time scales are considered. When stress induced compaction (consol...... (consolidation) is retarded by slow drainage of excess pore pressure it is expected that consolidation and creep occur simultaneously. A constitutive model adressing the problems of rate sensitive behavior and simultaneous consolidation and creep is presented....

  6. Ultrasonic irradiation preparation of graphitic-C3N4/polyaniline nanocomposites as counter electrodes for dye-sensitized solar cells.

    Science.gov (United States)

    Afshari, Mohaddeseh; Dinari, Mohammad; Momeni, Mohamad Mohsen

    2018-04-01

    In this research, polyaniline/graphitic carbon nitride (PANI/g-C 3 N 4 ) nanocomposites were synthesized via in-situ electrochemical polymerization of aniline monomer whit different number of cyclic voltammetry scans (10, 20 and 30 cycles) after electrode surface pre-preparation using a potential shock under ultrasonic irradiation. PANI/g-C 3 N 4 nanocomposites with two values of g-C 3 N 4 (0.010 wt% and 0.015 wt%) were deposited on the surface of the transparent conducting film (FTO glass) by immersing FTO into the aniline solution and g-C 3 N 4 during the electro-polymerization. The resulting PANI/g-C 3 N 4 films were characterized by Fourier transformed infra-red (FTIR), power X-ray diffraction (PXRD), field emission scanning electron microscopy (FE-SEM) and transmission electron microscopy (TEM) techniques. The prepared electrodes were applied as counter electrode in dye-sensitized solar cells. Among them, the prepared electrode with 10 cycles and 0.01 wt% g-C 3 N 4 showed the best efficiency. These hybrids show good catalytic activity in elevating tri-iodide reduction and due to the synergistic effect of PANI and g-C 3 N 4 , PANI/g-C 3 N 4 nanocomposite electrode shows power conversion efficiency about 1.8%. Copyright © 2017 Elsevier B.V. All rights reserved.

  7. Graphitic carbon nitride (g-C3N4)-Pt-TiO2 nanocomposite as an efficient photocatalyst for hydrogen production under visible light irradiation.

    Science.gov (United States)

    Chai, Bo; Peng, Tianyou; Mao, Jing; Li, Kan; Zan, Ling

    2012-12-28

    Porous graphitic carbon nitride (g-C(3)N(4)) was prepared by a simple pyrolysis of urea, and then a g-C(3)N(4)-Pt-TiO(2) nanocomposite was fabricated via a facile chemical adsorption followed by a calcination process. The obtained products were characterized by X-ray diffraction, X-ray photoelectron spectroscopy, UV-vis diffuse reflectance absorption spectra, and electron microscopy. It is found that the visible-light-induced photocatalytic hydrogen evolution rate can be remarkably enhanced by coupling TiO(2) with the above g-C(3)N(4), and the g-C(3)N(4)-Pt-TiO(2) composite with a mass ratio of 70 : 30 has the maximum photoactivity and excellent photostability for hydrogen production under visible-light irradiation, and the stable photocurrent of g-C(3)N(4)-TiO(2) is about 1.5 times higher than that of the bare g-C(3)N(4). The above experimental results show that the photogenerated electrons of g-C(3)N(4) can directionally migrate to Pt-TiO(2) due to the close interfacial connections and the synergistic effect existing between Pt-TiO(2) and g-C(3)N(4) where photogenerated electrons and holes are efficiently separated in space, which is beneficial for retarding the charge recombination and improving the photoactivity.

  8. AGC-2 Irradiation Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf

  9. Sugarcane juice derived carbon dot–graphitic carbon nitride composites for bisphenol A degradation under sunlight irradiation

    Science.gov (United States)

    Wong, Jing Lin; Hak, Chen Hong; Tai, Jun Yan; Leong, Kah Hon; Saravanan, Pichiah

    2018-01-01

    Carbon dots (CDs) and graphitic carbon nitride (g-C3N4) composites (CD/g-C3N4) were successfully synthesized by a hydrothermal method using urea and sugarcane juice as starting materials. The chemical composition, morphological structure and optical properties of the composites and CDs were characterized using various spectroscopic techniques as well as transmission electron microscopy. X-ray photoelectron spectroscopy (XPS) results revealed new signals for carbonyl and carboxyl groups originating from the CDs in CD/g-C3N4 composites while X-ray diffraction (XRD) results showed distortion of the host matrix after incorporating CDs into g-C3N4. Both analyses signified the interaction between g-C3N4 and CDs. The photoluminescence (PL) analysis indicated that the presence of too many CDs will create trap states at the CD/g-C3N4 interface, decelerating the electron (e−) transport. However, the CD/g-C3N4(0.5) composite with the highest coverage of CDs still achieved the best bisphenol A (BPA) degradation rate at 3.87 times higher than that of g-C3N4. Hence, the charge separation efficiency should not be one of the main factors responsible for the enhancement of the photocatalytic activity of CD/g-C3N4. Instead, the light absorption capability was the dominant factor since the photoreactivity correlated well with the ultraviolet–visible diffuse reflectance spectra (UV–vis DRS) results. Although the CDs did not display upconversion photoluminescence (UCPL) properties, the π-conjugated CDs served as a photosensitizer (like organic dyes) to sensitize g-C3N4 and injected electrons to the conduction band (CB) of g-C3N4, resulting in the extended absorption spectrum from the visible to the near-infrared (NIR) region. This extended spectral absorption allows for the generation of more electrons for the enhancement of BPA degradation. It was determined that the reactive radical species responsible for the photocatalytic activity were the superoxide anion radical (O2

  10. Sugarcane juice derived carbon dot-graphitic carbon nitride composites for bisphenol A degradation under sunlight irradiation.

    Science.gov (United States)

    Sim, Lan Ching; Wong, Jing Lin; Hak, Chen Hong; Tai, Jun Yan; Leong, Kah Hon; Saravanan, Pichiah

    2018-01-01

    Carbon dots (CDs) and graphitic carbon nitride (g-C 3 N 4 ) composites (CD/g-C 3 N 4 ) were successfully synthesized by a hydrothermal method using urea and sugarcane juice as starting materials. The chemical composition, morphological structure and optical properties of the composites and CDs were characterized using various spectroscopic techniques as well as transmission electron microscopy. X-ray photoelectron spectroscopy (XPS) results revealed new signals for carbonyl and carboxyl groups originating from the CDs in CD/g-C 3 N 4 composites while X-ray diffraction (XRD) results showed distortion of the host matrix after incorporating CDs into g-C 3 N 4 . Both analyses signified the interaction between g-C 3 N 4 and CDs. The photoluminescence (PL) analysis indicated that the presence of too many CDs will create trap states at the CD/g-C 3 N 4 interface, decelerating the electron (e - ) transport. However, the CD/g-C 3 N 4 (0.5) composite with the highest coverage of CDs still achieved the best bisphenol A (BPA) degradation rate at 3.87 times higher than that of g-C 3 N 4 . Hence, the charge separation efficiency should not be one of the main factors responsible for the enhancement of the photocatalytic activity of CD/g-C 3 N 4 . Instead, the light absorption capability was the dominant factor since the photoreactivity correlated well with the ultraviolet-visible diffuse reflectance spectra (UV-vis DRS) results. Although the CDs did not display upconversion photoluminescence (UCPL) properties, the π-conjugated CDs served as a photosensitizer (like organic dyes) to sensitize g-C 3 N 4 and injected electrons to the conduction band (CB) of g-C 3 N 4 , resulting in the extended absorption spectrum from the visible to the near-infrared (NIR) region. This extended spectral absorption allows for the generation of more electrons for the enhancement of BPA degradation. It was determined that the reactive radical species responsible for the photocatalytic activity were

  11. Creep in ceramics

    CERN Document Server

    Pelleg, Joshua

    2017-01-01

    This textbook is one of its kind, since there are no other books on Creep in Ceramics. The book consist of two parts: A and B. In part A general knowledge of creep in ceramics is considered, while part B specifies creep in technologically important ceramics. Part B covers creep in oxide ceramics, carnides and nitrides. While covering all relevant information regarding raw materials and characterization of creep in ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

  12. Effects of composition on the in-reactor creep of AISI 316

    International Nuclear Information System (INIS)

    Bates, J.F.; Gilbert, E.R.

    1979-08-01

    In-reactor tests designed to provide information on the relationship between compositional variations and irradiation-induced swelling and creep have achieved an exposure of 4.6 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 0 C. Postirradiation diametral measurements of pressurized tube specimens have indicated that irradiation-induced creep of 316 stainless steel can be modified by compositional variations of minor alloying elements. There is a general trend for specimens with higher swelling to exhibit higher creep. Silicon, phosphorus and molybdenum all retard in-reactor creep and inhibit irradiation-induced swelling as well. However, the relationship between creep and swelling is strongly composition dependent. The data suggest that carbon and nitrogen act synergistically the major influence being the nitrogen concentration. The irradiation-induced creep is insensitive to cobalt variations to the fluences investigated

  13. Effects of composition on the in-reactor creep of AISI 316

    International Nuclear Information System (INIS)

    Bates, J.F.; Gilbert, E.R.

    1980-01-01

    Pre- and postirradiation measurements of pressurized tube specimens irradiated at 450/degree/C to 4.6*10/sup 22/ n/cm/sup 2/(E>0.1 MeV) have indicated that increases in the solute concentrations of silicon, phosphorus, and molybdenum retard irradiation creep. The data suggest that carbon and nitrogen act synergistically with the major influence on creep being the nitrogen concentration. Irradiation-induced creep is insensitive to cobalt variations. There is a trend for specimens with higher swelling to exhibit higher creep. As the shear modulus increases, irradiation creep also increases. This shear modulus correlation is opposite to one observed for thermal creep deformation. 8 refs

  14. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  15. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Liu, Y.Y.; Arthur, B.

    1984-11-01

    The tritium breeding material with the highest lithium atom density, Li 2 O has been observed to incur significant swelling (>4%) under fast reactor irradiation. Such swelling, if unrestrained leads to either unacceptable, induced-strains in adjacent structural material or undesirable design compromises. Fortunately, however, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory dilational creep experiments were conducted on unirradiated Li 2 O between 500 and 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  16. Nanoindentation creep versus bulk compressive creep of dental resin-composites.

    Science.gov (United States)

    El-Safty, S; Silikas, N; Akhtar, R; Watts, D C

    2012-11-01

    To evaluate nanoindentation as an experimental tool for characterizing the viscoelastic time-dependent creep of resin-composites and to compare the resulting parameters with those obtained by bulk compressive creep. Ten dental resin-composites: five conventional, three bulk-fill and two flowable were investigated using both nanoindentation creep and bulk compressive creep methods. For nano creep, disc specimens (15mm×2mm) were prepared from each material by first injecting the resin-composite paste into metallic molds. Specimens were irradiated from top and bottom surfaces in multiple overlapping points to ensure optimal polymerization using a visible light curing unit with output irradiance of 650mW/cm(2). Specimens then were mounted in 3cm diameter phenolic ring forms and embedded in a self-curing polystyrene resin. Following grinding and polishing, specimens were stored in distilled water at 37°C for 24h. Using an Agilent Technologies XP nanoindenter equipped with a Berkovich diamond tip (100nm radius), the nano creep was measured at a maximum load of 10mN and the creep recovery was determined when each specimen was unloaded to 1mN. For bulk compressive creep, stainless steel split molds (4mm×6mm) were used to prepare cylindrical specimens which were thoroughly irradiated at 650mW/cm(2) from multiple directions and stored in distilled water at 37°C for 24h. Specimens were loaded (20MPa) for 2h and unloaded for 2h. One-way ANOVA, Levene's test for homogeneity of variance and the Bonferroni post hoc test (all at p≤0.05), plus regression plots, were used for statistical analysis. Dependent on the type of resin-composite material and the loading/unloading parameters, nanoindentation creep ranged from 29.58nm to 90.99nm and permanent set ranged from 8.96nm to 30.65nm. Bulk compressive creep ranged from 0.47% to 1.24% and permanent set ranged from 0.09% to 0.38%. There was a significant (p=0.001) strong positive non-linear correlation (r(2)=0.97) between bulk

  17. Mechanical properties of graphite and carbon materials

    International Nuclear Information System (INIS)

    Jouquet, G.

    1976-01-01

    The elastic properties of the graphite monocrystal, the role of internal characteristics (texture, porosity) on the mechanical behavior of carbons, effects caused by the gaseous environment and neutron irradiation, and the resistance of graphites to cyclic mechanical stresses are discussed [fr

  18. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  19. Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Li, M.; Stubbins, J.F.

    2005-01-01

    results of these investigations are presented andtheir implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the caseof...... causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K, on the other hand, causes anincrease in the number of cycles to failure. The irradiation at 573 K to a dose level of 0.2-0.3 dpa does...

  20. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil [Sogang Univ., Seoul, (Korea, Republic of); Lee, Jin Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-11-15

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve.

  1. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    International Nuclear Information System (INIS)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil; Lee, Jin Haeng

    2013-01-01

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve

  2. Creep properties of discontinuous fibre composites with partly creeping fibres

    International Nuclear Information System (INIS)

    Bilde-Soerensen, J.B.; Lilholt, H.

    1977-05-01

    In a previous report (RISO-M-1810) the creep properties of discontinuous fibre composites with non-creeping fibres were analyzed. In the present report this analysis is extended to include the case of discontinuous composites with partly creeping fibres. It is shown that the creep properties of the composite at a given strain rate, epsilonsub(c), depend on the creep properties of the matrix at a strain rate higher than epsilonsub(c), and on the creep properties of the fibres at epsilonsub(c). The composite creep law is presented in a form which permits a graphical determination of the composite creep curve. This can be constructed on the basis of the matrix and the fibre creep curves by vector operations in a log epsilon vs. log sigma diagram. The matrix contribution to the creep strength can be evaluated by a simple method. (author)

  3. Experience with graphite in JET

    International Nuclear Information System (INIS)

    Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

    1987-01-01

    During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

  4. Physical mechanisms of radiation induced creep in metals

    International Nuclear Information System (INIS)

    Borodin, V.A.; Ryazanov, A.I.

    1990-01-01

    Analysis of available experimental data has been conducted. It enables to correlate reliably the character of evolution of dislocation structure of irradiated materials with different stages of radiation induced creep. This provides reliable basis for the general conclusions concerning the character of some parametric dependences of deformation rate of these materials. Analysis of different modern theoretical models enables to evaluate regions of their applicability and their relative significance for radiation induced creep description. 20 refs.; 2 figs.; 1 tab

  5. ANSYS Creep-Fatigue Assessment tool for EUROFER97 components

    Directory of Open Access Journals (Sweden)

    M. Mahler

    2016-12-01

    Full Text Available The damage caused by creep-fatigue is an important factor for materials at high temperatures. For in-vessel components of fusion reactors the material EUROFER97 is a candidate for structural application where it is subjected to irradiation and cyclic thermo-mechanical loads. To be able to evaluate fusion reactor components reliably, creep-fatigue damage has to be taken into account. In the frame of Engineering Data and Design Integration (EDDI in EUROfusion Technology Work Programme rapid and easy design evaluation is very important to predict the critical regions under typical fusion reactor loading conditions. The presented Creep-Fatigue Assessment (CFA tool is based on the creep-fatigue rules in ASME Boiler Pressure Vessel Code (BPVC Section 3 Division 1 Subsection NH which was adapted to the material EUROFER97 and developed for ANSYS. The CFA tool uses the local stress, maximum elastic strain range and temperature from the elastic analysis of the component performed with ANSYS. For the assessment design fatigue and stress to rupture curves of EUROFER97 as well as isochronous stress vs. strain curves determined by a constitutive model considering irradiation influence are used to deal with creep-fatigue damage. As a result allowable number of cycles based on creep-fatigue damage interaction under given hold times and irradiation rates is obtained. This tool can be coupled with ANSYS MAPDL and ANSYS Workbench utilizing MAPDL script files.

  6. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  7. The optical properties and photocatalytic activity of CdS-ZnS-TiO{sub 2}/Graphite for isopropanol degradation under visible light irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rahmawati, Fitria, E-mail: fitria@mipa.uns.ac.id; Wulandari, Rini, E-mail: riniwulandari55@yahoo.com; Murni, Irvinna M., E-mail: irvinna-mutiara@yahoo.com; Mudjijono, E-mail: mbahparto@yahoo.com [Research Group of Solid State Chemistry & Catalysis, Chemistry Department, Sebelas Maret University, Jl. Ir. Sutami 36 A Kentingan, Surakarta, 57126 (Indonesia)

    2016-02-08

    This research prepared a photocatalyst tablet of CdS-ZnS-TiO{sub 2} on a graphite substrate. The synthesis was conducted through chemical bath deposition method. The graphite substrate used was a waste graphite rod from primary batteries. The aims of this research are studying the crystal structure, the optical properties and the photocatalytic activity of the prepared material. The photocatalytic activity was determined through isopropanol degradation. The result shows that the TiO{sub 2}/Graphite provide direct transition gap energy at 2.91 eV and an indirect transition gap energy at 3.21 eV. Deposition of CdS-ZnS changed the direct transition gap energy to 3.01 eV and the indirect transition gap energy to 3.22 eV. Isopropanol degradation with the prepared catalyst produced new peaks at 223-224 nm and 265-266 nm confirming the production of acetone. The degradation follows first order with rate constant of 2.4 × 10{sup −2} min{sup −1}.

  8. Biaxial Creep Specimen Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  9. Biaxial Creep Specimen Fabrication

    International Nuclear Information System (INIS)

    JL Bump; RF Luther

    2006-01-01

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

  10. Seismic Creep, USA Images

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Seismic creep is the constant or periodic movement on a fault as contrasted with the sudden rupture associated with an earthquake. It is a usually slow deformation...

  11. Limit analysis via creep

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.

    1981-07-01

    In this paper it is presented a variational method for the limit analysis of an ideal plastic solid. This method has been denominated as Modified Secundary Creep and enables to find the collapse loads through a minimization of a functional and a limit process. Given an ideal plastic material it is shown how to determinate the associated secundary creep constitutive equation. Finally, as an application, it is found the limit load in an pressurized von Mises rigid plastic sphere. (Author) [pt

  12. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hull, Laurence C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Technology Development Office

    2014-08-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC-3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC-3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. The report documents qualification of AGC-3 experiment irradiation monitoring data following MCP-2691. This report also documents whether AGC-3 experiment irradiation monitoring data meet the requirements for data collection as specified in technical and functional requirements documents and quality assurance (QA) plans. Data handling is described showing how data are passed from the data collection experiment to the Nuclear Data Management and Analysis System (NDMAS) team. The data structure is described, including data batches, components, attributes, and response variables. The description of the approach to data qualification includes the steps taken to qualify the data and the specific tests used to verify that the data meet requirements. Finally, the current status of the data received by NDMAS from the AGC-3 experiment is presented with summarized information on test results and resolutions. This report addresses all of the irradiation monitoring data collected during the AGC-3 experiment.

  13. Topical Problems and Applications of Creep Theory

    Science.gov (United States)

    Altenbach, H.

    2003-06-01

    A historical review of achievements in creep theory is given. Primary attention is focused on the phenomenological approach. Different constitutive equations are discussed for primary and secondary creep as well as for creep with damage. New creep problems are examined

  14. Creep in generation IV nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Rissanen, L. (VTT Technical Research Centre of Finland, Espoo (Finland))

    2010-05-15

    Nuclear power has an important role in fulfilling the world's growing energy needs and reducing the carbon dioxide emission. Six new, innovative nuclear energy systems have been identified and selected for further development by the international Generation Four International Forum (GIF). These generation four (Gen IV) nuclear energy systems include a variety of reactor, energy conversion and fuel cycle technologies. The successful development and deployment of these largely depend on the performance and reliability of the available structural materials. These potential materials need to sustain their mechanical properties up to high temperatures, high neutron doses and corrosive environments of the new or enhanced types of coolants. Current knowledge on material properties, material-coolant interaction and especially material degradation processes in these new environments are limited. This paper gives an overview of the Gen IV material issues with special emphasis on European design of supercritical light water reactor concept high performance light water reactor (HPLWR). The challenges for the structural materials and the components most likely to suffer from creep and creep-irradiation are highlighted. Some results from relatively short term creep testing in supercritical water are presented for AISI 316NG, 347H and 1.4970 steels. The 1.4970 steel was superior in creep and oxidation resistance (orig.)

  15. Radiation behaviour of graphite for HTGR

    International Nuclear Information System (INIS)

    Shtrombakh, Ya.I.; Platonov, P.A.; Gurovich, B.A.; Alekseev, V.M.

    1996-01-01

    The paper presents the results of investigations of different graphite materials, among with the standard reactor graphite manufacturing by electrode technology and a number of advanced graphites of new generation. During the investigation of radiation stability of standard reactor graphite the basic mechanisms of radiation damage of its structure were studied. With the help of transmission electron microscopy deformations and cracking of filler and binder were detected in the vicinity of the boundaries, separating these two components. Cracking begins with crystallite splitting and ends in full fracture of boundary layers. Such process of degradation can be explained by disjoint deformations resulting from difference in growth rate of filler and binder crystallites, in its turn caused by considerable difference between their sizes. It has been concluded that radiation stability of graphite may be improved by creating such graphite materials, in which the difference in sizes of crystallites of different structure components would be the minimal possible. When developing production technology of isotropic graphite for high temperature reactors, some progress was made towards the solution of this problem. Despite considerable swelling at high temperature this type of graphite appeared to be substantially less susceptible to the degradation of the structure and to deterioration of physico-mechanical properties. In addition to graphites manufactured by tradition technology, the graphite was investigated, in which pyrocarbon precipitated from gas phase under 1000 deg. C was used as binder. Carbon precipitated in such a way was non-graphitized at high temperatures and therefore it demonstrated sharp shrinkage under irradiation at high temperature, and shrinkage rate correlated with pyrocarbon quota in graphite structure. (author). 5 refs, 18 figs, 1 tab

  16. Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

    Directory of Open Access Journals (Sweden)

    Bespala Evgeny

    2016-01-01

    Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

  17. Predicting creep strengths and lifetimes of creep resistant engineering alloys

    Science.gov (United States)

    Zhao, Yanrong; Yao, Hongpeng; Song, Xinli; Jia, Juan; Xiang, Zhidong

    2018-01-01

    The physical basis for predicting the long-term creep strengths and lifetimes at application temperatures using creep parameters determined from short-term creep tests is investigated for complex creep resistant engineering alloys. It is shown that the seemingly unpredictable stress and temperature dependence of minimum creep rate of such alloys can be rationalised using an approach based on the new power law creep equation that incorporate the tensile strength. This is demonstrated using the tensile and creep data measured for two completely different types of alloys: steel 11Cr-2W-0.4Mo-1Cu-Nb-V and Ni base superalloy 15Cr-28Co-4Mo-2.5Ti-3Al. For both alloys, the stress exponent n determined does not depend on temperature and activation energy of creep does not depend on stress. Consequently, it becomes possible to use the new power law creep equation in combination with the Monkman-Grant relationship to predict the long term creep rupture strengths and lifetimes and microstructure stability of the two alloys from short term creep test data. The implications of the results for creep mechanism identification and future microstructure analysis are discussed.

  18. Elastic properties of graphite and interstitial defects

    International Nuclear Information System (INIS)

    Ayasse, J.-B.

    1977-01-01

    The graphite elastic constants C 33 and C 44 , reflecting the interaction of the graphitic planes, were experimentally measured as a function of irradiation and temperature. A model of non-central strength atomic interaction was established to explain the experimental results obtained. This model is valid at zero temperature. The temperature dependence of the elastic properties was analyzed. The influence of the elastic property variations on the specific heat of the lattice at very low temperature was investigated [fr

  19. Creep of fibrous composite materials

    DEFF Research Database (Denmark)

    Lilholt, Hans

    1985-01-01

    Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions to the cr......Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions...... such as Ni + W-fibres, high temperature materials such as Ni + Ni3Al + Cr3C2-fibres, and medium temperature materials such as Al + SiC-fibres. For the first two systems reasonable consistency is found for the models and the experiments, while for the third system too many unquantified parameters exist...

  20. AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Hull, Laurence Charles

    2016-01-01

    The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing

  1. AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Hull, Laurence Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing

  2. Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

    International Nuclear Information System (INIS)

    Atsumo, H.

    1977-01-01

    The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

  3. Micro creep mechanisms of tungsten

    International Nuclear Information System (INIS)

    Levoy, R.; Hugon, I.; Burlet, H.; Baillin, X.; Guetaz, L.

    2000-01-01

    Due to its high melting point (3410 deg C), tungsten offers good mechanical properties at elevated temperatures for several applications in non-oxidizing environment. The creep behavior of tungsten is well known between 1200 and 2500 deg C and 10 -3 to 10 -1 strain. However, in some applications when dimensional stability of components is required, these strains are excessive and it is necessary to know the creep behavior of the material for micro-strains (between 10 -4 and 10 -6 ). Methods and devices used to measure creep micro-strains are presented, and creep equations (Norton and Chaboche laws) were developed for wrought, annealed and recrystallized tungsten. The main results obtained on tungsten under low stresses are: stress exponent 1, symmetry of micro-strains in creep-tension and creep-compression, inverse creep (threshold stress), etc. TEM, SEM and EBSD studies allow interpretation of the micro-creep mechanism of tungsten under low stresses and low temperature (∼0.3 K) like the Harper-Dorn creep. In Harper-Dorn creep, micro-strains are associated with the density and the distribution of dislocations existing in the crystals before creep. At 975 deg C, the initial dislocation structure moves differently whether or not a stress is applied. To improve the micro-creep behavior of tungsten, a heat treatment is proposed to create the optimum dislocation structure. (authors)

  4. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Chow, C.K.

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  5. Consistent creep and rupture properties for creep-fatigue evaluation

    International Nuclear Information System (INIS)

    Schultz, C.C.

    1978-01-01

    The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a mininum strength heat is also shown to provide adequate predictions. The viability of using consistent properties (either actual or those of a minimum heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations

  6. Analysis of indentation creep

    Science.gov (United States)

    Don S. Stone; Joseph E. Jakes; Jonathan Puthoff; Abdelmageed A. Elmustafa

    2010-01-01

    Finite element analysis is used to simulate cone indentation creep in materials across a wide range of hardness, strain rate sensitivity, and work-hardening exponent. Modeling reveals that the commonly held assumption of the hardness strain rate sensitivity (mΗ) equaling the flow stress strain rate sensitivity (mσ...

  7. Graphite target for the spiral project

    Energy Technology Data Exchange (ETDEWEB)

    Putaux, J.C.; Ducourtieux, M.; Ferro, A.; Foury, P.; Kotfila, L.; Mueller, A.C.; Obert, J.; Pauwels, N.; Potier, J.C.; Proust, J. [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire; Bertrand, P. [Grand Accelerateur National d`Ions Lourds (GANIL), 14 - Caen (France); Loiselet, M. [Universite Catholique de Louvain, Louvain-La-Neuve (Belgium)] [and others

    1996-12-31

    A study of the thermal and physical properties of graphite targets for the SPIRAL project is presented. The main objective is to develop an optimized set-up both mechanically and thermally resistant, presenting good release properties (hot targets with thin slices). The results of irradiation tests concerning the mechanical and thermal resistance of the first prototype of SPIRAL target with conical geometry are presented. The micro-structural properties of the graphite target is also studied, in order to check that the release properties are not deteriorated by the irradiation. Finally, the results concerning the latest pilot target internally heated by an electrical current are shown. (author). 5 refs.

  8. Graphite reactor physics

    International Nuclear Information System (INIS)

    Bacher, P.; Cogne, F.

    1964-01-01

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm 2 , channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [fr

  9. Creep behavior of submarine sediments

    Science.gov (United States)

    Silva, Armand J.; Booth, J.S.

    1984-01-01

    A series of experiments on drained creep of marine sediment indicates that strength degradation results from the creep process, which implies an associated reduction in slope stability. Furthermore, the highest creep potential of a sediment may be at its preconsolidation stress. Results from the experiments on samples from Georges Bank continental slope were also used in conjunction with a preliminary theoretical model to predict creep displacements. For the case illustrated in this report, steep slopes (>20??) and thick sections (>30 m) give rise to substantial creep and probable creep rupture; as angles or thicknesses decrease, displacements rapidly become negligible. Creep may be a significant geologic process on many marine slopes. Not only can it cause major displacements of surface sediment, but it may also be the precursor to numerous slope failures. ?? 1985 Springer-Verlag New York Inc.

  10. Creep behaviour and creep mechanisms of normal and healing ligaments

    Science.gov (United States)

    Thornton, Gail Marilyn

    Patients with knee ligament injuries often undergo ligament reconstructions to restore joint stability and, potentially, abate osteoarthritis. Careful literature review suggests that in 10% to 40% of these patients the graft tissue "stretches out". Some graft elongation is likely due to creep (increased elongation of tissue under repeated or sustained load). Quantifying creep behaviour and identifying creep mechanisms in both normal and healing ligaments is important for finding clinically relevant means to prevent creep. Ligament creep was accurately predicted using a novel yet simple structural model that incorporated both collagen fibre recruitment and fibre creep. Using the inverse stress relaxation function to model fibre creep in conjunction with fibre recruitment produced a superior prediction of ligament creep than that obtained from the inverse stress relaxation function alone. This implied mechanistic role of fibre recruitment during creep was supported using a new approach to quantify crimp patterns at stresses in the toe region (increasing stiffness) and linear region (constant stiffness) of the stress-strain curve. Ligament creep was relatively insensitive to increases in stress in the toe region; however, creep strain increased significantly when tested at the linear region stress. Concomitantly, fibre recruitment was evident at the toe region stresses; however, recruitment was limited at the linear region stress. Elevating the water content of normal ligament using phosphate buffered saline increased the creep response. Therefore, both water content and fibre recruitment are important mechanistic factors involved in creep of normal ligaments. Ligament scars had inferior creep behaviour compared to normal ligaments even after 14 weeks. In addition to inferior collagen properties affecting fibre recruitment and increased water content, increased glycosaminoglycan content and flaws in scar tissue were implicated as potential mechanisms of scar creep

  11. Effect of graphite reflector on activation of fusion breeding blanket

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong Won; Cho, Seungyon; Ahn, Mu-Young

    2016-01-01

    Highlights: • The graphite reflector concept has been applied in the design of the Korea HCCR TBM for ITER and this concept is also a candidate design option for Korea Demo. • In the graphite reflector, C-14, B-11 and Be-10 are produced after an irradiation. Impurities in both case of beryllium and graphite is dominant in the shutdown dose after an irradiation. • Based on the evaluation, the graphite reflector is a good alternative of the beryllium multiplier in the view of induced activity and shutdown dose. But C-14 produced in the graphite reflector should be considered carefully in the view of radwaste management. - Abstract: Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. In this paper, activity analysis was performed and the effect of graphite reflector in the view of activation was compared to the beryllium multiplier. As a result, it is expected that using the graphite reflector instead of the beryllium multiplier decreases total activity very effectively. But the graphite reflector produces C-14 about 17.2 times than the beryllium multiplier. Therefore, C-14 produced in the graphite reflector is expected as a significant nuclide in the view of radwaste management.

  12. Theoretical basis for graphite stress analysis in BERSAFE

    International Nuclear Information System (INIS)

    Harper, P.G.

    1980-03-01

    The BERSAFE finite element computer program for structural analysis has been extended to deal with structures made from irradiated graphite. This report describes the material behaviour which has been modelled and gives the theoretical basis for the solution procedure. (author)

  13. Reassembling Surveillance Creep

    DEFF Research Database (Denmark)

    Bøge, Ask Risom; Lauritsen, Peter

    2017-01-01

    We live in societies in which surveillance technologies are constantly introduced, are transformed, and spread to new practices for new purposes. How and why does this happen? In other words, why does surveillance “creep”? This question has received little attention either in theoretical...... development or in empirical analyses. Accordingly, this article contributes to this special issue on the usefulness of Actor-Network Theory (ANT) by suggesting that ANT can advance our understanding of ‘surveillance creep’. Based on ANT’s model of translation and a historical study of the Danish DNA database......, we argue that surveillance creep involves reassembling the relations in surveillance networks between heterogeneous actors such as the watchers, the watched, laws, and technologies. Second, surveillance creeps only when these heterogeneous actors are adequately interested and aligned. However...

  14. Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

  15. Analysis of Picosecond Pulsed Laser Melted Graphite

    Science.gov (United States)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M. S.; Huang, C. Y.; Malvezzi, A. M.; Bloembergen, N.

    1986-12-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm{sup -1} and the disorder-induced mode at 1360 cm{sup -1}, the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nanosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence.

  16. Analysis of picosecond pulsed laser melted graphite

    Energy Technology Data Exchange (ETDEWEB)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm/sup -1/ and the disorder-induced mode at 1360 cm/sup -1/, the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence.

  17. Analysis of picosecond pulsed laser melted graphite

    International Nuclear Information System (INIS)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm -1 and the disorder-induced mode at 1360 cm -1 , the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence

  18. Remaining life assessment of carbon steel boiler headers by repeated creep testing

    Energy Technology Data Exchange (ETDEWEB)

    Drew, M. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia)]. E-mail: michael.drew@ansto.gov.au; Humphries, S. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Thorogood, K. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Barnett, N. [BlueScope Steel, P.O. Box 1854, Wollongong, NSW (Australia)

    2006-05-15

    The condition of carbon steel boiler headers that have been in service for over 25 years has been assessed periodically by NDT, dimensional measurements, replication and accelerated creep testing. Historical temperature records were limited, so estimates of effective header temperatures were made from replicas. These estimates were compared with header stub thermocouple readings. At about 280,000 service hours, samples were chain-drilled from the headers for accelerated creep testing. These test results indicated that the headers had satisfactory remaining life. Nine years after the original samples were taken, additional samples were removed from one header at 337,000 service hours. The creep rupture properties measured from the repeated tests were almost identical to the initial results. A mild degree of random, nodular graphite was found in the samples and its effect on creep properties is discussed.

  19. Creep in sodium

    International Nuclear Information System (INIS)

    Charnock, W.; Cordwell, J.E.

    1978-03-01

    Available information on the creep of austenitic, ferritic and Alloy-800 type steels in liquid sodium is critically reviewed. Creep properties of stainless steels can be affected by element transfer and corrosion. At reactor structural component temperatures environmental effects are likely to be less important than changes due to thermal ageing. At high clad temperatures (700 0 C) decarburisation may cause the loss of strength and ductility in unstabilised steels while cavity formation may cause embrittlement in stabilised steels. The properties of Alloy 800 are, in some experiments, found to deteriorate while in others they are enhanced. This may be a consequence of the metallurgical complexity of the material or arise from the nature of the various techniques employed. Low alloy ferritic steels tend to decarburise in sodium at temperatures greater than 500 0 C and this leads to loss of strength and an increase in ductility. High alloy ferritics are immune to this effect and appear to be able to tolerate a degree of carburisation. Although intergranular cracking may be enhanced in liquid sodium the mechanical consequences are not significant and evidence for the existence of an embrittlement effect not associated with element transfer or corrosion is weak. Stress and strain may enhance element transfer at crack tips. However in real cracks the gettering or supply action of the crack faces conditions the chemistry of the cracks in sodium and protects the crack tip from element transfer. Thus creep crack extension rates should be independent of changes in bulk coolant chemistry. (author)

  20. Creep of fibrous composite materials

    DEFF Research Database (Denmark)

    Lilholt, Hans

    1985-01-01

    Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions...... to the creep strength of composites. The advantage of combined analyses of several data sets is emphasized and illustrated for some experimental data. The analyses show that it is possible to derive creep equations for the (in situ) properties of the fibres. The experiments treated include model systems...... such as Ni + W-fibres, high temperature materials such as Ni + Ni3Al + Cr3C2-fibres, and medium temperature materials such as Al + SiC-fibres. For the first two systems reasonable consistency is found for the models and the experiments, while for the third system too many unquantified parameters exist...

  1. Numerical algorithms in secondary creep

    International Nuclear Information System (INIS)

    Feijoo, R.A.; Taroco, E.

    1980-01-01

    The problem of stationary creep is presented as well as its variational formulation, when weak constraints are established, capable of assuring one single solution. A second, so-called elasto-creep problem, is further analysed, together with its variational formulation. It is shown that its stationary solution coincides with that of the stationary creep and the advantages of this formulation with respect to the former one is emphasized. Some numerical applications showing the efficiency of the method propesed are finally presented [pt

  2. Effect of gamma radiation on graphite - PTFE dry lubrication system

    Science.gov (United States)

    Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

    2017-12-01

    An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties.

  3. On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

    The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

  4. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-01-01

    Several proposals have been made to assist adesigners with thermal ratcheting in the creep range, the more known has been made by O'DONNELL and POROWSKY. Unfortunately these methods are not validated by experiments, and they take only inelastic distortion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies - in providing an experimental basis to ratcheting analysis rules in the creep range, - in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimens made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture. (orig./GL)

  5. Creep failure of a spray drier

    CSIR Research Space (South Africa)

    Carter, P

    1998-06-01

    Full Text Available , and creep. The calculations pointed to creep, and no positive metallurgic or physical evidence was discovered to support any of the hypotheses. However, the compression stresses implied that creep deformation could have occurred without inducing discernible...

  6. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    International Nuclear Information System (INIS)

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs

  7. Tensile cracks in creeping solids

    International Nuclear Information System (INIS)

    Riedel, H.; Rice, J.R.

    1979-02-01

    The loading parameter determining the stress and strain fields near a crack tip, and thereby the growth of the crack, under creep conditions is discussed. Relevant loading parameters considered are the stress intensity factor K/sub I/, the path-independent integral C*, and the net section stress sigma/sub net/. The material behavior is modelled as elastic-nonlinear viscous where the nonlinear term describes power law creep. At the time t = 0 load is applied to the cracked specimen, and in the first instant the stress distribution is elastic. Subsequently, creep deformation relaxes the initial stress concentration at the crack tip, and creep strains develop rapidly near the crack tip. These processes may be analytically described by self-similar solutions for short times t. Small scale yielding may be defined. In creep problems, this means that elastic strains dominate almost everywhere except in a small creep zone which grows around the crack tip. If crack growth ensues while the creep zone is still small compared with the crack length and the specimen size, the stress intensity factor governs crack growth behavior. If the calculated creep zone becomes larger than the specimen size, the stresses become finally time-independent and the elastic strain rates can be neglected. In this case, the stress field is the same as in the fully-plastic limit of power law hardening plasticity. The loading parameter which determines the near tip fields uniquely is then the path-independent integral C*.K/sub I/ and C* characterize opposite limiting cases. The case applied in a given situation is decided by comparing the creep zone size with the specimen size and the crack length. Besides several methods of estimating the creep zone size, a convenient expression for a characteristic time is derived, which characterizes the transition from small scale yielding to extensive creep of the whole specimen

  8. Vortex pinning and creep experiments

    International Nuclear Information System (INIS)

    Kes, P.H.

    1991-01-01

    A brief review of basic flux-pinning and flux-creep ingredients and a selection of experimental results on high-temperature-superconductivity compounds is presented. Emphasis is put on recent results and on those properties which are central to the emerging understanding of the flux-pinning and flux-creep mechanisms of these fascinating materials

  9. Nanogranular origin of concrete creep.

    Science.gov (United States)

    Vandamme, Matthieu; Ulm, Franz-Josef

    2009-06-30

    Concrete, the solid that forms at room temperature from mixing Portland cement with water, sand, and aggregates, suffers from time-dependent deformation under load. This creep occurs at a rate that deteriorates the durability and truncates the lifespan of concrete structures. However, despite decades of research, the origin of concrete creep remains unknown. Here, we measure the in situ creep behavior of calcium-silicate-hydrates (C-S-H), the nano-meter sized particles that form the fundamental building block of Portland cement concrete. We show that C-S-H exhibits a logarithmic creep that depends only on the packing of 3 structurally distinct but compositionally similar C-S-H forms: low density, high density, ultra-high density. We demonstrate that the creep rate ( approximately 1/t) is likely due to the rearrangement of nanoscale particles around limit packing densities following the free-volume dynamics theory of granular physics. These findings could lead to a new basis for nanoengineering concrete materials and structures with minimal creep rates monitored by packing density distributions of nanoscale particles, and predicted by nanoscale creep measurements in some minute time, which are as exact as macroscopic creep tests carried out over years.

  10. Creep buckling analysis of shells

    International Nuclear Information System (INIS)

    Stone, C.M.; Nickell, R.E.

    1977-01-01

    The current study was conducted in an effort to determine the degree of conservatism or lack of conservatism in current ASME design rules concerning time-dependent (creep) buckling. In the course of this investigation, certain observations were made concerning the numerical solution of creep buckling problems. It was demonstrated that a nonlinear finite element code could be used to solve the time-dependent buckling problem. A direct method of solution was presented which proved to be computationally efficient and provided answers which agreed very well with available analytical solutions. It was observed that the calculated buckling times could vary widely for small errors in computed displacements. The presence of high creep strain rates contributed to the prediction of early buckling times when calculated during the primary creep stage. The predicted time estimates were found to increase with time until the secondary stage was reached and the estimates approached the critical times predicted without primary creep. It can be concluded, therefore, that for most nuclear piping components, whose primary creep stage is small compared to the secondary stage, the effect of primary creep is negligible and can be omitted from the calculations. In an evaluation of the past and current ASME design rules for time-dependent, load controlled buckling, it was concluded that current use of design load safety factors is not equivalent to a safety factor of ten on service life for low creep exponents

  11. IFMIF - Design Study for in Situ Creep Fatigue Tests

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Simakov, St.; Stratmanns, E.; Vladimirov, P.; Moeslang, A.

    2006-01-01

    While the high flux volume (20-50 dpa/fpy) of the International Fusion Materials Irradiation Facility (IFMIF) is dedicated to the irradiation of ∼ 1100 qualified specimens that will be post irradiation examined after disassembling in dedicated Hot Cells, various in situ experiments are foreseen in the medium flux volume (1-20 dpa/fpy). Of specific importance for structural lifetime assessments of fusion power reactors are instrumented in situ creep-fatigue experiments, as they can simulate realistically a superposition of thermal fatigue or creep fatigue and irradiation with fusion relevant neutrons. Based on former experience with in situ fatigue tests under high energy light ion irradiation, a design study has been performed to evaluate the feasibility of in situ creep fatigue tests in the IFMIF medium flux position. The vertically arranged test module for such experiments consists basically of a frame similar to a universal testing machine, but equipped with three pulling rods, driven by independent step motors, instrumentation systems and specimen cooling systems. Therefore, three creep fatigue specimens may be tested at one time in this apparatus. Each specimen is a hollow tube with coolant flow in the specimen interior to maintain individual specimen temperatures. The recently established IFMIF global 3D geometry model was used together the latest McDeLicious code for the neutral and charged particle transport calculations. These comprehensive neutronics calculations have been performed with a fine special resolution of 0.25 cm 3 , showing among others that the specimens will be irradiated with a homogeneous damage rate of up to 13(∼ 9%) dpa/fpy and a fusion relevant damage to helium ratio of 10-12 appm He/dpa. In addition, damage and gas production rates as well as the heat deposition in structural parts of the test module have been calculated. Despite of the vertical gradients in the nuclear heating, CFD code calculations with STAR-CD revealed very

  12. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-08-01

    Creep is a cause of deformation; it may also result in rupture in time. Although LMFBR structures are not heavily loaded, they are subjected to large thermal transients. Can structure lifetime be shortened by such transients. Several proposals have been made to assist adesigners with thermal ratcheting in the creep range. Unfortunately these methods are not validated by experiments, and they take only inelastic distorsion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies in providing an experimental basis to ratcheting analysis rules in the creep range, and in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimen made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture

  13. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  14. Temperature-dependence of creep behaviour of dental resin-composites.

    Science.gov (United States)

    El-Safty, S; Silikas, N; Watts, D C

    2013-04-01

    To determine the effect of temperature, over a clinically relevant range, on the creep behaviour of a set of conventional and flowable resin-composites including two subgroups having the same resin matrix and varied filler loading. Eight dental resin-composites: four flowable and four conventional were investigated. Stainless steel split moulds (4 mm × 6 mm) were used to prepare cylindrical specimens for creep examination. Specimens were irradiated in the moulds in layers of 2mm thickness (40s each), as well as from the radial direction after removal from the moulds, using a light-curing unit with irradiance of 650 mW/cm(2). A total of 15 specimens from each material were prepared and divided into three groups (n=5) according to the temperature; Group I: (23°C), Group II: (37°C) and Group III: (45°C). Each specimen was loaded (20 MPa) for 2h and unloaded for 2h. Creep was measured continuously over the loading and unloading periods. At higher temperatures greater creep and permanent set were recorded. The lowest mean creep occurred with GS and GH resin-composites. Percentage of creep recovery decreased at higher temperatures. At 23°C, the materials exhibited comparable creep. At 37°C and 45°C, however, there was a greater variation between materials. For all resin-composites, there was a strong linear correlation with temperature for both creep and permanent set. Creep parameters of resin-composites are sensitive to temperature increase from 23 to 45°C, as can occur intra-orally. For a given resin matrix, creep decreased with higher filler loading. Copyright © 2012 Elsevier Ltd. All rights reserved.

  15. Diamond amorphization in neutron irradiation

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Gordeev, V.G.

    1996-01-01

    The paper presents the results on neutron irradiation of the diamond in a nuclear reactor. It is shown that the neutron irradiation stimulates the diamond transition to the amorphous state. At a temperature below 750 o K the time required for the diamond-graphite transition decreases with decreasing irradiation temperature. On the contrary, in irradiation at higher temperatures the time of diamond conversion into the amorphous state increases with decreasing but always remains shorter than in the absence of irradiation. (author)

  16. Contribution to the influence of selected alloy elements on the strain cycling and creep behaviour of cast iron with spheroidal graphite at temperatures above 450 C; Beitrag zum Einfluss ausgewaehlter Legierungselemente auf das Dehnwechsel- und Zeitstandverhalten von Gusseisen mit Kugelgraphit bei Temperaturen oberhalb 450 C

    Energy Technology Data Exchange (ETDEWEB)

    Michel, Susanne

    2012-02-15

    In this report for the first time an all-embracing databases was raised about the correlation of microstructure and high temperature mechanical properties concerning spheroidal cast iron at temperatures above 450 C. Its basic concept is a systematic variation of alloying elements and benchmarking fatigue and creep behavior of all created heats as a function of microstructure and alloys.

  17. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

  18. Mechanisms of transient radiation-induced creep

    International Nuclear Information System (INIS)

    Pyatiletov, Yu.S.

    1981-01-01

    Radiation-induced creep at the transient stage is investigated for metals. The situation, when several possible creep mechanisms operate simultaneously is studied. Among them revealed are those which give the main contribution and determine thereby the creep behaviour. The time dependence of creep rate and its relation to the smelling rate is obtained. The results satisfactorily agree with the available experimental data [ru

  19. Negative creep in nickel base superalloys

    DEFF Research Database (Denmark)

    Dahl, Kristian Vinter; Hald, John

    2004-01-01

    Negative creep describes the time dependent contraction of a material as opposed to the elongation seen for a material experiencing normal creep behavior. Negative creep occurs because of solid state transformations that results in lattice contractions. For most applications negative creep will h...

  20. Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

    Science.gov (United States)

    Hall, M. M., Jr.; Flinn, J. E.

    2010-01-01

    Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.

  1. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence Hull

    2014-10-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC 3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC 3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. All thermocouples (TCs) functioned throughout the AGC 3 experiment. There was one interval between December 18, 2012, and December 20, 2012, where 10 NULL values were reported for various TCs. These NULL values were deleted from the Nuclear Data Management and Analysis System database. All temperature data are Qualified for use by the VHTR TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the VHTR TDO Program. Total gas flow was approximately 50 sccm through the AGC 3 experiment capsule. Helium gas flow was briefly increased to 100 sccm during ATR shutdowns. At the start of the AGC 3 experiment, moisture in the outflow gas line was stuck at a constant value of 335.6174 ppmv for the first cycle (Cycle 152B). When the AGC 3 experiment capsule was reinstalled in ATR for Cycle 154B, a new moisture filter was installed. Moisture data from Cycle 152B are Failed. All moisture data from the final three cycles (Cycles 154B, 155A, and 155B) are Qualified for use by the VHTR TDO Program.

  2. Multiaxial creep-fatigue rules

    International Nuclear Information System (INIS)

    Spindler, M.W.; Hales, R.; Ainsworth, R.A.

    1997-01-01

    Within the UK, a comprehensive procedure, called R5, is used to assess the high temperature response of structures. One part of R5 deals with creep-fatigue initiation, and in this paper we describe developments in this part of R5 to cover multiaxial stress states. To assess creep-fatigue, damage is written as the linear sum of fatigue and creep components. Fatigue is assessed using Miner's law with the total endurance split into initiation and growth cycles. Initiation is assessed by entering the curve of initiation cycles vs strain range using a Tresca equivalent strain range. Growth is assessed by entering the curve of growth cycles vs strain range using a Rankine equivalent strain range. The number of allowable cycles is obtained by summing the initiation and growth cycles. In this way the problem of defining an equivalent strain range applicable over a range of endurance is avoided. Creep damage is calculated using ductility exhaustion methods. In this paper we address two aspects; first, the nature of stress relaxation and, hence, accumulated creep strain in multiaxial stress fields; secondly, the effect of multiaxial stress on creep ductility. The effect of multiaxial stress state on creep ductility has been examined using experimental data and mechanistic models. Good agreement is demonstrated between an empirical description of test data and a cavity growth model, provided a simple nucleation criterion is included. A simple scaling factor is applied to uniaxial creep ductility, defined as a function of stress state. The factor is independent of the cavity growth mechanisms and yields a value of equivalent strain which can be conveniently used in determining creep damage by ductility exhaustion. (author). 14 refs, 4 figs

  3. Topological investigation of nuclear graphite using small angle scattering

    Science.gov (United States)

    Rai, Durgesh K.; Khaykovich, Boris; Campbell, Anne A.; Ilvasky, Jan; Katoh, Yutai; Snead, Lance L.

    Nuclear power reactors require high performance materials that withstand high temperatures and neutron damage over long period of times. Graphite is widely used for high temperature fission reactor applications. It has a complex multiphase microstructure, which is affected by neutron irradiation. The irradiation-induced microstructures result in significant thermophysical property changes, affecting service lifetimes. It is important to understand these life-limiting phenomena at many different length scales. We present the results from small angle scattering (SAS) studies on graphite samples, which vary in doses and irradiation temperatures. The neutron and synchrotron SAS measurement data indicates that the graphite morphology consists of surface fractal structures. The samples were found to be uniform across several decades of length scale, while exhibiting different surface fractal dimensions, for different irradiation doses and temperature conditions. The surface fractal dimension changes at HFIR at ORNL, DOE User Facility; APS at ANL, DOE User Facility; Office of Nuclear Energy NSUF.

  4. Creep in electronic ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Routbort, J. L.; Goretta, K. C.; Arellano-Lopez, A. R.

    2000-04-27

    High-temperature creep measurements combined with microstructural investigations can be used to elucidate deformation mechanisms that can be related to the diffusion kinetics and defect chemistry of the minority species. This paper will review the theoretical basis for this correlation and illustrate it with examples from some important electronic ceramics having a perovskite structure. Recent results on BaTiO{sub 3}, (La{sub 1{minus}x}Sr){sub 1{minus}y}MnO{sub 3+{delta}}, YBa{sub 2}Cu{sub 3}O{sub x}, Bi{sub 2}Sr{sub 2}CaCu{sub 2}O{sub x}, (Bi,Pb){sub 2}Sr{sub 2}Ca{sub 2}Cu{sub 3}O{sub x} and Sr(Fe,Co){sub 1.5}O{sub x} will be presented.

  5. Mechanical behaviour of nuclear fuel under irradiation

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    The main mechanical properties (fracture, thermal and irradiation creep) of oxide and carbide fuels are summarised and discussed. Some examples are given of the influence of these mechanical properties on the in-pile behaviour of fuel pins [fr

  6. Numerical description of creep of highly creep resistant alloys

    International Nuclear Information System (INIS)

    Preussler, T.

    1991-01-01

    Fatigue tests have been performed with a series of highly creep resistant materials for gas turbines and related applications for gaining better creep data up to long-term behaviour. The investigations were performed with selected individual materials in the area of the main applications down to strains and stresses relevant to design, and have attained trial durations of 25000 to 60000 h. In continuing former research, creep equations for a selection of characterizing individual materials have been improved and partly newly developed on the basis of a differentiated evaluation. Concerning the single materials, there are: one melt each of the materials IN-738 LC, IN-939, IN-100, FSX-414 and Inconel 617. The applied differentiated evaluation is based on the elastoplastical behaviour from the hot-drawing test, the creep behaviour from the non interrupted or the interrupted fatigue test, and the contraction behaviour from the annealing test. The creep equations developed describe the high temperature deformation behaviour taking into account primary, secondary and partly the tertiary creep dependent of temperature, stress and time. These equations are valid for the whole application area of the respective material. (orig./MM) [de

  7. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  8. Facile one-pot synthesis of cerium oxide/sulfur-doped graphitic carbon nitride (g-C3N4) as efficient nanophotocatalysts under visible light irradiation.

    Science.gov (United States)

    Jourshabani, Milad; Shariatinia, Zahra; Badiei, Alireza

    2017-12-01

    Porous CeO 2 /sulfur-doped g-C 3 N 4 (CeO 2 /CNS) composites were synthesized by one-pot thermal condensation of thiourea and cerium nitrate as starting materials. The obtained CeO 2 (x)/CNS composites (x=8.4, 9.5 and 10.4wt%) with different CeO 2 contents were characterized by the XRD, FT-IR, XPS, TEM, BET, DRS and PL analyses. The TEM images displayed a nonporous and platelet-like morphology for pure CNS but a nanoporous structure with numerous uniform pore sizes of ∼40nm for the CeO 2 (9.5)/CNS composite. The XRD phase structures and TEM morphologies confirmed that structural evolution trend and stacking degree of CNS were disrupted in precense of the CeO 2 nanoparticles. The optimized photocatalyst, i.e. CeO 2 (9.5)/CNS nanocomposite, exhibited the highest visible light photocatalytic activity (91.4% after 150min) with a reaction rate constant of 0.0152min -1 toward methylene blue (MB) degradation which was greater compared with the individual CNS (0.0044min -1 ) and CeO 2 (0.0031min -1 ) photocatalysts. This enhanced photocatalytic performance was originated from heterojunctions formed between CeO 2 and CNS that improved the effective charge transfer through interfacial interactions between both components. The heterojunction prepared displayed excellent stability for the photocatalytic activity under the optimized conditions including catalyst dosage 0.08g, initial dye concentration 7mg/L and irradiation time 150min which was obtained using response surface methodology (RSM). The trapping experiments using isopropanol, benzoquinone and ethylenediaminetetraacetic as the OH, O 2 - and h + scavengers, respectively, verified that the OH and O 2 - as major species directly attacked onto the MB molecules while h + showed a negligible role. Finally, it could be stated that simultaneous doping of both sulfur and CeO 2 within the g-C 3 N 4 structure using a simple one-pot synthetic process produced very active photocatalysts illustrating their potential for

  9. Plasticity and creep of metals

    CERN Document Server

    Rusinko, Andrew

    2011-01-01

    Here is a systematic presentation of the postulates, theorems and principles of mathematical theories of plasticity and creep in metals, and their applications. Special attention is paid to analysis of the advantages and shortcomings of the classical theories.

  10. Creep in an electrodeposited nickel

    Czech Academy of Sciences Publication Activity Database

    Sklenička, Václav; Kuchařová, Květa; Kvapilová, Marie; Svoboda, Milan; Král, Petr; Vidrich, G.

    2013-01-01

    Roč. 48, č. 13 (2013), s. 4780-4788 ISSN 0022-2461 R&D Projects: GA ČR(CZ) GAP108/11/2260; GA MŠk(CZ) ED1.1.00/02.0068 Institutional support: RVO:68081723 Keywords : nanocrystalline nickel * electrodeposited metals * nanocomposite * creep behavior * creep mechanisms Subject RIV: JG - Metallurgy Impact factor: 2.305, year: 2013

  11. Creep deformation behaviour and microstructural changes in Zr-2.5% Nb alloy

    International Nuclear Information System (INIS)

    Chaudhuri, S.; Singh, R.; Ghosh, R.N.; Sinha, T.K.; Banerjee, S.

    2002-01-01

    Cold worked and stress relieved Zr-2.5% Nb alloy is a well-known material used as pressure tubes in Pressurised Heavy Water Reactors. The pressure tubes, made of a typical Zr-alloy, consisting of 2.54% Nb, 0.1175% oxygen and less than 100 ppm impurities, are expected to withstand 9.5 MPa to 12.5 MPa pressure at 250 degC to 310 degC under fast neutron fluxes of 3.5 x 10 17 nm -2 s -1 . These tubes are made by hot extrusion at 780 degC with an extrusion ratio 8.3:1 and 40% cold pilgering followed by annealing at 550 degC for 3 hours and subsequently by 20-30% cold pilgering and stress relieving at 400 degC for 24 hours. The microstructure of such cold worked and stress relieved alloy consists of Β-Zr precipitates in the matrix of elongated Α-Zr grains. Although various factors such as irradiation creep, thermal creep, irradiation growth etc are responsible for limiting the life of pressure tubes; the thermal creep contributes significantly in overall creep deformation. Keeping this in view as well as due to non-availability of adequate published information including creep database on this alloy, an extensive investigation on the thermal creep behaviour of indigenously produced Zr-2.5% Nb alloy was undertaken. The creep tests in air using Mayes' creep testing machines were carried out in the temperature range of 300 degC to 450 degC under stresses in the range of 50 to 550 MPa. Analysis of data revealed that the mechanism of creep deformation remains the same in this range

  12. Critical survey of the neutron-induced creep behaviour of steel alloys for the fusion reactor materials programme

    International Nuclear Information System (INIS)

    Hausen, H.

    1985-01-01

    The differences between the irradiation environment of a fission reactor and that of a fusion reactor are respectively described in relation to the radiation damage found and expected in the two types of nuclear reactor. It is shown that the microstructure developing for instance in stainless steel alloys is almost invariant to whether the production rate of helium is high or low. The finding is valid up to neutron doses corresponding to about 60 dpa. For this reason, irradiation creep data obtained in fission reactors may be used, with caution, for predicting creep behaviour in fusion reactors.It was further recognized that irradiation creep performed with high energy particles from an accelerator, yields results which are comparable to those obtained in fission reactors. For this reason, simulation creep experiments are found to be valuable for the development of irradiation creep resistant materials using, for example, high energy electrons or protons. Such kind of experiments are performed in many laboratories. For irradiation doses larger than 60 dpa, predictions with respect to creep rates in fission and fusion reactors are difficult. In end-of-life tests, which concern swelling, ductility, tensile properties, rupture, fatigue and embrittlement, the presence of helium, due to its production rate being much higher in most materials exposed to 14 MeV neutrons than to fission neutrons, may be of great importance

  13. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  14. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...

  15. Graphitic Carbon Nanocubes Derived from ZIF-8 for Photothermal Therapy.

    Science.gov (United States)

    Chen, Wei; Zhang, Xiaoman; Ai, Fujin; Yang, Xueqing; Zhu, Guangyu; Wang, Feng

    2016-06-20

    Graphitic carbon nanocubes (GCNCs) were prepared by pyrolysis of ZIF-8 nanocubes. The GCNCs resemble the structure of N-doped graphite and exhibit a high photothermal conversion efficiency of 40.4%. In vitro tests demonstrate that the GCNCs are highly biocompatible and induce an effective photothermal therapy effect under 808 nm irradiation. Our study provides a facile strategy for preparing functional carbon nanomaterials of prescribed size, morphology, and porous structure for bioapplications.

  16. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

  17. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  18. Oxidizability and explosibility of pure graphite powder

    International Nuclear Information System (INIS)

    L Rahmani; D Roubineau; S Cornet

    2005-01-01

    Full text of publication follows: While graphite is widely considered a heat-resistant material, e.g. able to screen metallic shielding from thermal damage, and graphite powder is used as a fire extinguisher agent where water or carbon dioxide should not, it still can react with air and - being carbon - give forth a significant amount of heat. Whether this makes it a hazard in operations such as dismantling nuclear reactors that contain hundreds of tons of graphite, including a small percentage of powder, is a question that has to be answered, considering that dismantling implies the use of such potential fire initiators as thermal cutters and electrical equipment. For this reason EDF commissioned the Centre National de Prevention et Protection (CNPP) to carry out explosibility tests on unirradiated, nuclear grade (i.e. with about 100 ppm of impurities) graphite powder. CNPP tests were so designed as to simulate realistic conditions that might result from a severe mishap during a dismantling operation, such as the crash of heavy equipment on graphite blocks coupled with the bruise of a high power electrical cable. EDF-CNPP tests complement others, done either in Italy most notably on irradiated graphite dust contaminated with various pollutants, or in the UK where the ability of settled graphite dust to propagate an initial gas explosion into an adjacent volume was assessed. EDF-CNPP tests comprise two steps. Step one was intended to produce a qualitative understanding of how nuclear grade graphite behaves while heated in air. In a first series of experiments graphite samples were heated up to 900 C during two and a half hours and their mass loss measured: it was found that while fine powder is wholly oxidised, coarser powder and chunks retained about two thirds of their initial mass. Oxidation kinetics, as assessed by oven temperature shoot-up, begins at 580 C and is quite low, compared with that of iron powder. In a second series of experiments a graphite piece

  19. Oxidizability and explosibility of pure graphite powder

    International Nuclear Information System (INIS)

    Rahmani, L.; Roubineau, D.; Cornet, S.

    2005-01-01

    Full text of publication follows: While graphite is widely considered a heat-resistant material, e.g. able to screen metallic shielding from thermal damage, and graphite powder is used as a fire extinguisher agent where water or carbon dioxide should not, it still can react with air and - being carbon - give forth a significant amount of heat. Whether this makes it a hazard in operations such as dismantling nuclear reactors that contain hundreds of tons of graphite, including a small percentage of powder, is a question that has to be answered, considering that dismantling implies the use of such potential fire initiators as thermal cutters and electrical equipment. For this reason EDF commissioned the Centre National de Prevention et Protection (CNPP) to carry out explosibility tests on unirradiated, nuclear grade (i.e. with about 100 ppm of impurities) graphite powder. CNPP tests were so designed as to simulate realistic conditions that might result from a severe mishap during a dismantling operation, such as the crash of heavy equipment on graphite blocks coupled with the bruise of a high power electrical cable. EDF-CNPP tests complement others, done either in Italy most notably on irradiated graphite dust contaminated with various pollutants, or in the UK where the ability of settled graphite dust to propagate an initial gas explosion into an adjacent volume was assessed. EDF-CNPP tests comprise two steps. Step one was intended to produce a qualitative understanding of how nuclear grade graphite behaves while heated in air. In a first series of experiments graphite samples were heated up to 900 C during two and a half hours and their mass loss measured: it was found that while fine powder is wholly oxidised, coarser powder and chunks retained about two thirds of their initial mass. Oxidation kinetics, as assessed by oven temperature shoot-up, begins at 580 C and is quite low, compared with that of iron powder. In a second series of experiments a graphite piece

  20. Effect of temperature changes on swelling and creep of AISI 316

    International Nuclear Information System (INIS)

    Garner, F.A.; Gilbert, E.R.; Gelles, D.S.; Foster, J.P.

    1980-04-01

    A number of previous publications have shown that the swelling of cold-worked AISI 316 is quite sensitive to changes in temperature which occur during irradiation. In this report those data are expanded and reanalyzed to show that the concurrent irradiation creep is also quite sensitive to changes in irradiation temperature. An explanation is advanced to explain this behavior in terms of the sensitivity to temperture history of the radiation-induced microchemical evolution of this steel. In particular, the sensitivity to temperature history of the radiation-stabilized gamma prime phase is invoked to explain the enhanced creep and swelling behavior of AISI 316 components which experienced either gradual or abrupt decreases in temperature. The phase development observed in this steel in response to temperature changes during irradiation is also compared to the similar behavior found in aged specimens subjected to isothermal irradiation

  1. Datalogger for the creep laboratory

    International Nuclear Information System (INIS)

    Sambasivan, S.I.; Karthikeyan, T.V.; Chowdhary, D.M.; Anantharaman, P.N.

    1989-01-01

    The creep laboratory, MDL/ICGAR is a facility to study the creep properties of materials which are of interest to the fast reactor programme. The creep test is conducted over a few days to several months and years depending on the test variables employed. In these tests the creep strain and creep rate as a function of time are studied while the load and temperature are kept constant. The datalogger automates the process of recording the strain information as a function of time and also monitors the temperature throughout the test. The system handles 126 temperature channels and 42 strain channels from 27 machines. The temperature inputs are from the thermocouples and for cold junction compensation RTD's are used. An extensometer with a linear variable differential transformer (LVDT) or Super Linear Variable Capacitor (SLVC) form the set up to measure strain. The data logger consists of a front end analog input sub-system (AISS), a 8085 based Data Acquisition System (DAS) communicating to a microcomputer with CP/M operating system. The system responds to the user through the console and outputs of a dot matrix printer. The system, running a real time executive, also allows for on line enabling or disabling of a channel, printing of data, examining the current status and value, setting and getting time etc. (author)

  2. Creep of plasma sprayed zirconia

    Science.gov (United States)

    Firestone, R. F.; Logan, W. R.; Adams, J. W.

    1982-01-01

    Specimens of plasma-sprayed zirconia thermal barrier coatings with three different porosities and different initial particle sizes were deformed in compression at initial loads of 1000, 2000, and 3500 psi and temperatures of 1100 C, 1250 C, and 1400 C. The coatings were stabilized with lime, magnesia, and two different concentrations of yttria. Creep began as soon as the load was applied and continued at a constantly decreasing rate until the load was removed. Temperature and stabilization had a pronounced effect on creep rate. The creep rate for 20% Y2O3-80% ZrO2 was 1/3 to 1/2 that of 8% Y2O3-92% ZrO2. Both magnesia and calcia stabilized ZrO2 crept at a rate 5 to 10 times that of the 20% Y2O3 material. A near proportionality between creep rate and applied stress was observed. The rate controlling process appeared to be thermally activated, with an activation energy of approximately 100 cal/gm mole K. Creep deformation was due to cracking and particle sliding.

  3. Modelling deformation and fracture of Gilsocarbon graphite subject to service environments

    Science.gov (United States)

    Šavija, Branko; Smith, Gillian E.; Heard, Peter J.; Sarakinou, Eleni; Darnbrough, James E.; Hallam, Keith R.; Schlangen, Erik; Flewitt, Peter E. J.

    2018-02-01

    Commercial graphites are used for a wide range of applications. For example, Gilsocarbon graphite is used within the reactor core of advanced gas-cooled reactors (AGRs, UK) as a moderator. In service, the mechanical properties of the graphite are changed as a result of neutron irradiation induced defects and porosity arising from radiolytic oxidation. In this paper, we discuss measurements undertaken of mechanical properties at the micro-length-scale for virgin and irradiated graphite. These data provide the necessary inputs to an experimentally-informed model that predicts the deformation and fracture properties of Gilsocarbon graphite at the centimetre length-scale, which is commensurate with laboratory test specimen data. The model predictions provide an improved understanding of how the mechanical properties and fracture characteristics of this type of graphite change as a result of exposure to the reactor service environment.

  4. Analysis of radiation exposure during creep adjustment to the coolant channels at Madras Atomic Power Station

    International Nuclear Information System (INIS)

    Varadhan, R.S.; Venkataramana, K.; Kannan, R.K.; Sreekumaran Nair, B.; Chudalayandi, K.

    1994-01-01

    In pressurised heavy water reactors the coolant channels made of zircaloy-2 undergo creep deformation used intense neutron irradiation in the reactor core. In order to measure and provide for the changes in the dimensions, base line data of internal diameters, sag and length of the 306 coolant channels are measured as pre service inspection (PSI) before the reactor is loaded with fuel prior to criticality. Subsequently as part of in service inspection (ISI), axial creep of every channel is measured in every annual shutdown of the reactor and creep adjustment is done on those channels where creep expansion margin for the next one year operation is low. A study was carried out to assess the radiological impact of the job at Madras Atomic Power Station (MAPS). Various measures adopted for reducing the individual and collective doses on the job are discussed in this report. (author). 3 refs., 2 tabs

  5. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  6. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  7. Creep of service-aged welds

    OpenAIRE

    Sun, Wei

    1996-01-01

    The creep behaviour of welds in service-aged pipes are studied. The aims of the research have been achieved using analytical, numerical and experimental approaches to the relevant subjects. Several features of the work are presented: (i) a systematic parametric study of the creep of two-material test specimens including a stress singularity analysis, (ii) an impression creep testing method using a rectangular indenter, which can be applied to study the creep properties in welds, and (iii) met...

  8. Stress analysis of fuel claddings with axial fins including creep effects

    International Nuclear Information System (INIS)

    Krieg, R.

    1977-01-01

    For LMFBR fuel claddings with axial fins the stress and strain fields are calculated which may be caused by internal pressure, differential thermal expansion and irradiation induced differential swelling. To provide an appropriate description of the cladding material it is assumed that the total strain is the sum of a linear elastic and a creep term, where the latter one includes the thermal as well as the irradiation induced creep. First the linear elastic problem is treated by a semi-analytical method leading to a bipotential equation for Airys' stress function. Solving this equation analytically means that the field equations valid within the cladding are satisfied exactly. By applying a combined point matching- least square-method the boundary conditions could be satisfied approximately such that in most cases the remaining error is within the uncertainty range of the loading conditions. Then the nonlinear problem which includes creep is approximated by a sequence of linear elastic solutions with time as parameter. The accumulated creep strain is treated here as an imposed strain field. To study the influence of different effects such as fin shape, temperature region, irradiation induced creep and swelling or internal pressure, a total of eleven cases with various parameter variations are investigated. The results are presented graphically in the following forms: stress and strain distributions over the cladding cross section for end of life conditions and boundary stresses and strains versus time. (Auth.)

  9. Influence of variations in creep curve on creep behavior of a high-temperature structure

    International Nuclear Information System (INIS)

    Hada, Kazuhiko

    1986-01-01

    It is one of the key issues for a high-temperature structural design guideline to evaluate the influence of variations in creep curve on the creep behavior of a high-temperature structure. In the present paper, a comparative evaluation was made to clarify such influence. Additional consideration was given to the influence of the relationship between creep rupture life and minimum creep rate, i.e., the Monkman-Grant's relationship, on the creep damage evaluation. The consideration suggested that the Monkman-Grant's relationship be taken into account in evaluating the creep damage behavior, especially the creep damage variations. However, it was clarified that the application of the creep damage evaluation rule of ASME B and P.V. Code Case N-47 to the ''standard case'' which was predicted from the average creep property would predict the creep damage on the safe side. (orig./GL)

  10. Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

    International Nuclear Information System (INIS)

    Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

    2005-01-01

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

  11. Comparison of Interfacial Strengthening in Creep Deformation and Radiation Damage Processes of Advanced Structural Materials for Nuclear Applications

    Science.gov (United States)

    Zhu, Hanliang

    2018-02-01

    The mechanisms for microstructural strengthening in creep deformation and radiation damage processes of advanced structural materials for nuclear applications are compared. During creep and irradiation, various defects are generated and move in the microstructure. Any microstructural features that can retard such defect movement may improve both creep and radiation damage resistance. Interfaces in the microstructure are important barriers for preventing defect motion. To achieve ultrahigh strength and enhanced radiation damage resistance, an extremely high density of interfaces has been designed in recently developed nanostructured materials. However, interface-mediated processes may govern the deformation of these materials, decreasing their creep properties. Methods for improving the creep resistance of nanostructured materials are reviewed and discussed.

  12. Creep and Creep-Fatigue of Alloy 617 Weldments

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Jill K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carroll, Laura J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    Alloy 617 is the primary candidate material for the heat exchanger of a very high temperature gas cooled reactor intended to operate up to 950°C. While this alloy is currently qualified in the ASME Boiler and Pressure Vessel Code for non-nuclear construction, it is not currently allowed for use in nuclear designs. A draft Code Case to qualify Alloy 617 for nuclear pressure boundary applications was submitted in 1992, but was withdrawn prior to approval. Prior to withdrawal of the draft, comments were received indicating that there was insufficient knowledge of the creep and creep-fatigue behavior of Alloy 617 welds. In this report the results of recent experiments and analysis of the creep-rupture behavior of Alloy 617 welds prepared using the gas tungsten arc process with Alloy 617 filler wire. Low cycle fatigue and creep-fatigue properties of weldments are also discussed. The experiments cover a range of temperatures from 750 to 1000°C to support development of a new Code Case to qualify the material for elevated temperature nuclear design. Properties of the welded material are compared to results of extensive characterization of solution annealed plate base metal.

  13. SHMUTZ & PROTON-DIAMANT H + Irradiated/Written-Hyper/Super-conductivity(HC/SC) Precognizance/Early Experiments Connections: Wet-Graphite Room-Tc & Actualized MgB2 High-Tc: Connection to Mechanical Bulk-Moduli/Hardness: Diamond Hydrocarbon-Filaments, Disorder, Nano-Powders:C,Bi,TiB2,TiC

    Science.gov (United States)

    Wunderman, Irwin; Siegel, Edward Carl-Ludwig; Lewis, Thomas; Young, Frederic; Smith, Adolph; Dresschhoff-Zeller, Gieselle

    2013-03-01

    SHMUTZ: ``wet-graphite''Scheike-....[Adv.Mtls.(7/16/12)]hyper/super-SCHMUTZ-conductor(S!!!) = ``wet''(?)-``graphite''(?) = ``graphene''(?) = water(?) = hydrogen(?) =ultra-heavy proton-bands(???) = ...(???) claimed room/high-Tc/high-Jc superconductOR ``p''-``wave''/ BAND(!!!) superconductIVITY and actualized/ instantiated MgB2 high-Tc superconductors and their BCS- superconductivity: Tc Siegel[ICMAO(77);JMMM 7,190(78)] connection to SiegelJ.Nonxline-Sol.40,453(80)] disorder/amorphous-superconductivity in nano-powders mechanical bulk/shear(?)-moduli/hardness: proton-irradiated diamond, powders TiB2, TiC,{Siegel[Semis. & Insuls.5:39,47, 62 (79)])-...``VS''/concommitance with Siegel[Phys.Stat.Sol.(a)11,45(72)]-Dempsey [Phil.Mag. 8,86,285(63)]-Overhauser-(Little!!!)-Seitz-Smith-Zeller-Dreschoff-Antonoff-Young-...proton-``irradiated''/ implanted/ thermalized-in-(optimal: BOTH heat-capacity/heat-sink & insulator/maximal dielectric-constant) diamond: ``VS'' ``hambergite-borate-mineral transformable to Overhauser optimal-high-Tc-LiBD2 in Overhauser-(NW-periodic-table)-Land: CO2/CH4-ETERNAL-sequestration by-product: WATER!!!: physics lessons from

  14. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  15. Graphite Technology Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; T. Burchell; R. Bratton

    2007-09-01

    This technology development plan is designed to provide a clear understanding of the research and development direction necessary for the qualification of nuclear grade graphite for use within the Next Generation Nuclear Plant (NGNP) reactor. The NGNP will be a helium gas cooled Very High Temperature Reactor (VHTR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Considerable effort will be required to ensure that the graphite performance is not compromised during operation. Based upon the perceived requirements the major data needs are outlined and justified from the perspective of reactor design, reatcor performance, or the reactor safety case. The path forward for technology development can then be easily determined for each data need. How the data will be obtained and the inter-relationships between the experimental and modeling activities will define the technology development for graphite R&D. Finally, the variables affecting this R&D program are discussed from a general perspective. Factors that can significantly affect the R&D program such as funding, schedules, available resources, multiple reactor designs, and graphite acquisition are analyzed.

  16. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  17. Photooxidation Behavior of a LDPE/Clay Nanocomposite Monitored through Creep Measurements

    Directory of Open Access Journals (Sweden)

    Francesco Paolo La Mantia

    2017-07-01

    Full Text Available Creep behavior of polymer nanocomposites has not been extensively investigated so far, especially when its effects are combined with those due to photooxidation, which are usually studied in completely independent ways. In this work, the photooxidation behavior of a low density polyethylene/organomodified clay nanocomposite system was monitored by measuring the creep curves obtained while subjecting the sample to the combined action of temperature, tensile stress, and UV radiation. The creep curves of the irradiated samples were found to be lower than those of the non-irradiated ones and progressively diverging, because of the formation of branching and cross-linking due to photooxidation. This was further proved by the decrease of the melt index and the increase of the intrinsic viscosity; at the same time, the formation of carbonyl groups was observed. This behavior was more observable in the nanocomposite sample, because of its faster photooxidation kinetics.

  18. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    Energy Technology Data Exchange (ETDEWEB)

    P. Macheret

    2000-06-12

    The objective of this calculation is to compare several creep rupture correlations for use in calculating creep strain accrued by the Zircaloy cladding of spent nuclear fuel when it has been emplaced in the repository. These correlations are used to calculate creep strain values that are then compared to a large set of experimentally measured creep strain data, taken from four different research articles, making it possible to determine the best fitting correlation. The scope of the calculation extends to six different creep rupture correlations.

  19. Creep Resistance of VM12 Steel

    Directory of Open Access Journals (Sweden)

    Zieliński A.

    2016-09-01

    Full Text Available This article presents selected material characteristics of VM12 steel used for elements of boilers with super- and ultra-critical steam parameters. In particular, abridged and long-term creep tests with and without elongation measurement during testing and investigations of microstructural changes due to long-term impact of temperature and stress were carried out. The practical aspect of the use of creep test results in forecasting the durability of materials operating under creep conditions was presented. The characteristics of steels with regard to creep tests developed in this paper are used in assessment of changes in functional properties of the material of elements operating under creep conditions.

  20. Negative creep in nickel base superalloys

    DEFF Research Database (Denmark)

    Dahl, Kristian Vinter; Hald, John

    2004-01-01

    Negative creep describes the time dependent contraction of a material as opposed to the elongation seen for a material experiencing normal creep behavior. Negative creep occurs because of solid state transformations that results in lattice contractions. For most applications negative creep...... will have no practical implications but under certain conditions it may become critical. For bolts and fasteners, which are highly constrained during service, negative creep may lead to dramatically increased stresses and eventually to failure. The article was inspired by a recent failure of Nimonic 80A...

  1. Modelling of diffusional creep in polycrystals

    Energy Technology Data Exchange (ETDEWEB)

    Pein, Cornelia; Sommitsch, Christof [Technische Univ. Graz (Austria). Inst. for Materials Science and Welding

    2010-07-01

    To study creep behaviour on a microstructure level is of major importance, because the microstructure of metallic materials and its influence on creep phenomena is complex. Therefore a physically based finite element model is introduced to study the deformation behaviour due to diffusion creep phenomena. The influence of grain boundaries triple junctions and precipitates on creep strains and stresses is simulated. The results indicate that the different microstructure configurations, such as the presence of triple points, second phase particles and the relative orientation of grain boundaries to the loading direction influence the stress distribution and therefore lead to a highly heterogenous creep strain distribution. (orig.)

  2. Creep of ice: further studies

    International Nuclear Information System (INIS)

    Heard, H.C.; Durham, W.B.; Kirby, S.H.

    1987-01-01

    Detailed studies have been done of ice creep as related to the icy satellites, Ganymede and Callisto. Included were: (1) the flow of high-pressure water ices II, III, and V, and (2) frictional sliding of ice I sub h. Work was also begun on the study of the effects of impurities on the flow of ice. Test results are summarized

  3. Some equipment for graphite research in swimming pool reactors

    International Nuclear Information System (INIS)

    Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

    1964-01-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

  4. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Schulze, R.E.; Schulze, H.A.

    1981-02-01

    The report comprises the graphical documentation of irradiation results on graphitic matrix materials for spherical HTR fuel elements. The plotted results are based on data analyses of the series of exposures in the High Flux Reactor Petten (HFR). The documentation includes information about the changes of - the dimensions - the dynamic modulus of elasticity - the coefficient of thermal expansion of the materials after irradiation with fast neutrons. The irradiation experiments and the data analyses are part of the matrix development and irradiation programme, whose objective, realization and results obtained are summarized. (orig./IHOE) [de

  5. Low stress creep of stainless steel

    International Nuclear Information System (INIS)

    Crossland, I.G.; Clay, B.D.; Baker, C.

    1976-06-01

    The creep of 20%Cr, 25%Ni, Nb stainless steel has been examined at temperatures from 675 to 775 0 C at sheer stressed below 13 MPa and grain sizes from 6 to 20μm. The results have indicated that the initial creep rates were linearly dependent upon stress but with a threshold stress below which no creep occurred, i.e. Bingham behaviour; in addition, the creep activation energy at small strains was substantially lower than the lattice self-diffusion value and the initial creep rates were approximately related to the grain size through an inverse cube relation. It has been concluded that at low strains (approaching the initial elastic deflection) the creep mechanism was probably that of grain boundary diffusion creep (Coble, 1963) and this is further supported by the close agreement between the observed and theoretically predicted creep rate values. Steady-state creep rates were not observed; initially the creep rates fell rapidly with strain after which a more gradual decrease occurred. Whilst the creep rate - stress relationship continued to be of a Bingham form, the progressive reduction in creep rate with strain was found to be mainly attributable to an increase in the effective viscosity, threshold stress effects being generally of secondary importance. A model has been proposed which explains the initial creep rates as being due to Cable creep with elastic accommodation at grain boundary particles. At higher strains grain boundary collapse caused by vacancy sinking is accommodated at precipitate particles by plastic deformation of the adjacent matrix material. (author)

  6. On estimating the fracture probability of nuclear graphite components

    International Nuclear Information System (INIS)

    Srinivasan, Makuteswara

    2008-01-01

    The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation

  7. Influence of microstructure modification on the circumferential creep of Zr–Nb–Sn–Fe cladding tubes

    International Nuclear Information System (INIS)

    Jeong, Gu Beom; Kim, In Won; Hong, Sun Ig

    2016-01-01

    Out-of-reactor, non-irradiated thermal creep performances and lives of annealed and stress-relieved Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were studied and compared. The creep rates of annealed Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were appreciably slower than those of stress-relieved annealed counterpart. The stress exponent increased slightly from 5.1 to 6.1 in the stress-relieved cladding to 5.3–6.3 in the annealed cladding. The creep activation energy of the annealed Zr-1.02Nb-0.69Sn-0.12Fe alloy (300–330 kJ/mol) was larger compared to that of the stress-relieved alloy (210–260 kJ/mol). The creep activation energy of annealed alloy is close to that of self-diffusion in α-Zr (336 kJ/mol). The smaller activation energy in the stress-relieved alloy is attributed to the increasing contribution of faster diffusion path such as grain boundaries and dislocations. The presence of dislocation arrays with higher dislocation density and smaller grain size in the stress-relived alloy was confirmed by TEM analysis. The creep rupture time increased dramatically in the annealed Zr–1Nb- 0.7Sn-0.1Fe alloy compared to that of stress-relieved alloy, supporting the decrease of creep rate by annealing. The creep life of Zr-1.02Nb-0.69Sn-0.12Fe claddings can be extended through microstructure modification by annealing at intermediate temperatures in which dislocation creep dominates. - Highlights: • Effect of microstructure modification on creep in Zr–Nb–Sn–Fe tubes was studied. • Creep activation energy in annealed tubes was larger than in stress-relieved tubes. • Lower dislocation density in lager grains was observed after creep in annealed tubes. • Larson–Miller parameter of annealed tube was larger than that of stress-relieved one. • Creep life of tubes was extended through microstructure modification by annealing.

  8. Living with creep damage - outside the creep range

    Energy Technology Data Exchange (ETDEWEB)

    Brear, M.; Jarvis, P. [Stress Engineering Services, Europe, Ltd., Esher (United Kingdom)

    2007-06-15

    This paper addresses the effects of creep cavitation damage on other mechanical properties - chiefly those that affect behaviour outside the creep range. Such effects seem not to have been systematically studied, yet they are significant for the understanding and prediction of component integrity and life. The paper presents results obtained mainly as by-products of research programmes on low-alloy steels for both fossil and nuclear power plant and seeks to rationalise the findings to generate an overall picture of the effect. It is seen that a simple loss-of-effective-section model is adequate to describe many of the phenomena observed, but that other factors may also need consideration. (orig.)

  9. Stress Calculation of a TRISO Coated Particle Fuel by Using a Poisson's Ratio in Creep Condition

    International Nuclear Information System (INIS)

    Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Kim, W. K.

    2007-01-01

    KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) project since 2004, has been developing a performance analysis code for the TRISO coated particle fuel named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. KAERI, on the other hand, is developing an ABAQUS based finite element(FE) model to cover the non-linear behaviors of a coated particle fuel such as cracking or debonding of the TRISO coating layers. Using the ABAQUS based FE model, verification calculations were carried out for the IAEA CRP-6 benchmark problems involving creep, swelling, and pressure. However, in this model the Poisson's ratio for elastic solution was used for creep strain calculation. In this study, an improvement is made for the ABAQUS based finite element model by using the Poisson's ratio in creep condition for the calculation of the creep strain rate. As a direct input of the coefficient in a creep condition is impossible, a user subroutine for the ABAQUS solution is prepared in FORTRAN for use in the calculations of the creep strain of the coating layers in the radial and hoop directions of the spherical fuel. This paper shows the calculation results of a TRISO coated particle fuel subject to an irradiation condition assumed as in the Miller's publication in comparison with the results obtained from the old FE model used in the CRP-6 benchmark calculations

  10. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  11. Effect of microstructural evolution on in-reactor creep of Zr-2.5Nb tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, YoungSuk [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of)]. E-mail: yskim1@kaeri.re.kr; Im, KyungSoo [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of); Cheong, YongMoo [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of); Ahn, SangBok [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of)

    2005-11-15

    Dislocation density, decomposition of the {beta}-Zr phase and diametral creep were examined as a function of the location of the Zr-2.5Nb tube irradiated in the Wolsong Unit 1 for 9.3 effective full power years (EFPYs). The maximum a-dislocation density occurred at the inlet part of the irradiated Zr-2.5Nb tube exposed to the lowest temperature while the outlet part of the tube exposed to the higher temperature had the higher extent of decomposition of the {beta}-Zr phase and the maximum diametral creep. Thus, it is concluded that in-reactor creep of the Zr-2.5Nb tube is not related to the dislocation density but governed by the Nb concentration of the {alpha}-Zr grains caused by thermal decomposition of the {beta}-Zr phase. Supplementary creep tests on the Zr-2.5Nb sheets with different Nb contents in the {beta}-Zr phase provide supportive evidence to this conclusion. The acceleration of the in-reactor creep of the Zr-2.5Nb tubes is suggested after a long-term operation.

  12. Contribution to the study of internal friction in graphites

    International Nuclear Information System (INIS)

    Merlin, J.

    1969-03-01

    A study has been made of the internal friction in different graphites between -180 C and +500 C using a torsion pendulum; the graphites had been previously treated thermo-mechanically, by neutron irradiation and subjected to partial annealings. It has been shown that there occurs: a hysteretic type dissipation of energy, connected with interactions between dislocations and other defects in the matrix; a dissipation having a partially hysteretic character which can be interpreted by a Granato-Luke type formalism and which is connected with the presence of an 'ultra-micro porosity'; a dissipation by a relaxation mechanism after a small dose of irradiation; this is attributed to the reorientation of bi-interstitials; a dissipation having the characteristics of a solid state transformation, this during an annealing after irradiation. It is attributed to the reorganization of interstitial defects. Some information has thus been obtained concerning graphites, in particular: their behaviour at low mechanical stresses, the nature of irradiation defects and their behaviour during annealing, the structural changes occurring during graphitization, the relationship between internal friction and macroscopic mechanical properties. (author) [fr

  13. Cesium diffusion in graphite

    International Nuclear Information System (INIS)

    Evans, R.B. III; Davis, W. Jr.; Sutton, A.L. Jr.

    1980-05-01

    Experiments on diffusion of 137 Cs in five types of graphite were performed. The document provides a completion of the report that was started and includes a presentation of all of the diffusion data, previously unpublished. Except for data on mass transfer of 137 Cs in the Hawker-Siddeley graphite, analyses of experimental results were initiated but not completed. The mass transfer process of cesium in HS-1-1 graphite at 600 to 1000 0 C in a helium atmosphere is essentially pure diffusion wherein values of (E/epsilon) and ΔE of the equation D/epsilon = (D/epsilon) 0 exp [-ΔE/RT] are about 4 x 10 -2 cm 2 /s and 30 kcal/mole, respectively

  14. Structure and properties of crosslinked PTFE irradiated

    International Nuclear Information System (INIS)

    Kusano, Hiroo; Ikeda, Shigetoshi; Kasai, Noboru; Oshima, Akihiro; Seguchi, Tadao.

    1996-01-01

    Polytetrafluoroethylene (PTFE) was crosslinked by EB irradiation at the molten state in oxygen free atmosphere. The properties of crosslinked PTFE was investigated on the radiation resistance, the creep resistance and change of electric properties. The radiation resistance was much improved by the crosslinking, and the electric properties were not so much changed. The creep resistance at room temperature and at 200degC were also improved. (author)

  15. The influence of the grain boundary structure on diffusional creep

    International Nuclear Information System (INIS)

    Thorsen, P.A.

    1998-05-01

    An experiment was carried out to quantify the deformation in the diffusional creep domain. It was found that material had indisputably been deposited at grain boundaries in tension. A characterisation of 131 boundaries in terms of their misorientation was carried out and this was correlated to the observed deformation. Twin boundaries below a certain limit of deviation from an exact twin misorientation were totally inactive in the deformation. A large qualitative difference was found in the way general boundaries take part in the deformation. The experiments have taken place at Materials Research Department, Risoe National Laboratory at Roskilde. The present thesis has been submitted in partial fulfillment of the requirements for the Ph.D. degree in physics at the Niels Bohr Institute, University of Copenhagen. Besides the results of the creep experiment the thesis contains a description of the theoretical background to diffusional creep models. Also, the results from an investigation of helium bubble formation in an irradiated copper sample is included. (au)

  16. Creep in commercially pure metals

    International Nuclear Information System (INIS)

    Nabarro, F.R.N.

    2006-01-01

    The creep of commercially pure polycrystalline metals under constant stress has four stages: a virtually instantaneous extension, decelerating Andrade β creep, almost steady-state Andrade κ creep, and an acceleration towards failure. Little is known about the first stage, and the fourth stage has been extensively reviewed elsewhere. The limited experimental evidence on the physical mechanism of the second stage is reviewed and a critical discussion is given of various theories of this stage. The dependence of strain rate on stress in the third, steady-state, period seems to fall into two regimes, a power law with an exponent of about 4-5, and a rather closely exponential law. The limits of the parameters within which a simple theory of the exponential dependence can be expected to be valid are discussed, and found to be compatible with experiments. Theories of the power-law dependence are discussed, and, appear to be unconvincing. The theoretical models do not relate closely to the metallographic and other physical observations. In view of the weakness of theory, experiments which may indicate the physical processes dominant in steady-state creep are reviewed. It is usually not clear whether they pertain to the power-law or the exponential regime. While the theories all assume that most of the deformation occurs homogeneously within the grains, most experimental observations point strongly to a large deformation at or close to the grain boundaries. However, a detailed study of dislocation processes in a single grain of polycrystalline foil strained in the electron microscope shows that most of the observed strain can be accounted for by the motion of single dislocations through the subgrain structure. There is no clear reconciliation of these two sets of observations. Grain-boundary sliding cannot occur without intragranular deformation. One or other process may dominate the overall deformation; the geometrically dominant process may not be the rate

  17. Graphite-based photovoltaic cells

    Science.gov (United States)

    Lagally, Max; Liu, Feng

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

  18. A study on creep properties of laminated rubber bearings. Pt. 1. Creep properties and numerical simulations of thick rubber bearings

    International Nuclear Information System (INIS)

    Matsuda, Akihiro; Yabana, Shuichi

    2000-01-01

    In this report, to evaluate creep properties and effects of creep deformation on mechanical properties of thick rubber bearings for three-dimensional isolation system, we show results of compression creep test for rubber bearings of various rubber materials and shapes and development of numerical simulation method. Creep properties of thick rubber bearings were obtained from compression creep tests. The creep strain shows steady creep that have logarithmic relationships between strain and time and accelerated creep that have linear relationships. We make numerical model of a rubber material with nonlinear viscoelastic constitutional equations. Mechanical properties after creep loading test are simulated with enough accuracy. (author)

  19. Creep of parylene-C film

    KAUST Repository

    Lin, Jeffrey Chun-Hui

    2011-06-01

    The glass transition temperature of as-deposited parylene-C is first measured to be 50°C with a ramping-temperature-dependent modulus experiment. The creep behavior of parylene-C film in the primary and secondary creep region is then investigated below and above this glass transition temperature using a dynamic mechanical analysis (DMA) machine Q800 from TA instruments at 8 different temperatures: 10, 25, 40, 60, 80, 100, 120 and 150°C. The Burger\\'s model, which is the combined Maxwell model and Kelvin-Voigt model, fits well with our primary and secondary creep data. Accordingly, the results show that there\\'s little or no creep below the glass transition temperature. Above the glass transition temperature, the primary creep and creep rate increases with the temperature, with a retardation time constant around 6 minutes. © 2011 IEEE.

  20. Control of epoxy creep using graphene.

    Science.gov (United States)

    Zandiatashbar, Ardavan; Picu, Catalin R; Koratkar, Nikhil

    2012-06-11

    The creep behavior of epoxy-graphene platelet (GPL) nanocomposites with different weight fractions of filler is investigated by macroscopic testing and nanoindentation. No difference is observed at low stress and ambient temperature between neat epoxy and nanocomposites. At elevated stress and temperature the nanocomposite with the optimal weight fraction, 0.1 wt% GPLs, creeps significantly less than the unfilled polymer. This indicates that thermally activated processes controlling the creep rate are in part inhibited by the presence of GPLs. The phenomenon is qualitatively similar at the macroscale and in nanoindentation tests. The results are compared with the creep of epoxy-single-walled (SWNT) and multi-walled carbon nanotube (MWNT) composites and it is observed that creep in both these systems is similar to that in pure epoxy, that is, faster than creep in the epoxy-GPL system considered in this work. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Room temperature creep in metals and alloys

    Energy Technology Data Exchange (ETDEWEB)

    Deibler, Lisa Anne [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Characterization and Performance

    2014-09-01

    Time dependent deformation in the form of creep and stress relaxation is not often considered a factor when designing structural alloy parts for use at room temperature. However, creep and stress relaxation do occur at room temperature (0.09-0.21 Tm for alloys in this report) in structural alloys. This report will summarize the available literature on room temperature creep, present creep data collected on various structural alloys, and finally compare the acquired data to equations used in the literature to model creep behavior. Based on evidence from the literature and fitting of various equations, the mechanism which causes room temperature creep is found to include dislocation generation as well as exhaustion.

  2. In pile measurement of creep rate of stainless steel cladding tubes for fast reactor pins

    International Nuclear Information System (INIS)

    Calza Bini, A.; Cosoli, G.; Filacchioni, G.; Lanchi, M.; Nobili, A.; Pesce, E.; Rocca, U.V.; Rotoloni, P.L.

    1975-01-01

    Results are reported of a direct in pile measurement of creep on a cladding sample of 10cm length, under tensile stress of 22.82kg/mm 2 at a temperature of 550 0 during about 500 hours, up to an integrated flux of 2.6.10 20 n/cm 2 . Two identical samples were irradiated in the same temperature and flux conditions to be submitted to out of pile creep measurements together with other unirradiated samples. The aim of this first experiment was mainly to set up the device and to evaluate the kind and the quality of the available data

  3. In-Pile creep rupture properties of ODS ferritic steel claddings

    International Nuclear Information System (INIS)

    Kaito, T.; Uwaba, T.; Mizuta, S.; Ito, C.; Kagota, E.; Kitamura, R.; Ohtsuka, S.; Inoue, M.; Asayama, T.; Ukai, S.; Furukawa, T.; Inoue, T.

    2007-01-01

    Full text of publication follows: Oxide Dispersion Strengthened (ODS) ferritic steels are the most prospective material for both advanced sodium cooled fast breeder reactor (SFR) fuels and fusion reactor components. In the SFR core, superior radiation resistance and high temperature capability are essential for fuel pin cladding tubes which will be exposed to high neutron doses up to 250 dpa relevant to peak burnup of 250 GWd/t in high temperature flowing sodium ranging from 673 K to 973 K. Japan Atomic Energy Agency (JAEA) has been developing two types of ODS steels, which are 9Cr-ODS steel (9Cr-0.13C-2W-0.2Ti-0.35Y 2 O 3 ) and 12Cr-ODS steel (12Cr-0.05C-2W-0.3Ti-0.25Y 2 O 3 ). For the cladding tubes, internal creep rupture strength is one of the most important properties; for example, internal pressure gradually increases with burnup and finally reaches at 120 MPa in the highest burnup fuel pins. In order to examine irradiation effect on creep rupture strength of the ODS steels, an in-pile internal creep rupture test has been conducted in the experimental fast reactor JOYO using Material Testing Rig with Temperature Control (MARICO). Twenty-four pressurized tube specimens made from both 9Cr- and 12Cr-ODS steels have been irradiated at temperatures of 943 K, 973 K and 1023 K up to 20 dpa. Hoop stress for each specimen was varied with filling helium gas volume to attain predetermined pressure ranging from 45 MPa to 155 MPa at desired test temperature. Small amount of xenon and krypton mixed gas with unique isotopic composition was also filled into each specimen and released into cover gas systems after creep rupture in order to identify its creep rupture time by analyzing gas species by means of Laser Resonance Ionization Mass Spectrometry (RIMS). In MARICO test, 14 creep ruptures have been detected by the end of February 2007. Up to now, no irradiation effect on creep rupture strength of the ODS steels has been distinguished. This indicates that nanometer size

  4. Graphite technology development plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-07-01

    This document presents the plan for the graphite technology development required to support the design of the 350 MW(t) Modular HTGR within the US National Gas-Cooled Reactor Program. Besides descriptions of the required technology development, cost estimates, and schedules, the plan also includes the associated design functions and design requirements.

  5. Transitional Thermal Creep of Early Age Concrete

    DEFF Research Database (Denmark)

    Hauggaard-Nielsen, Anders Boe; Damkilde, Lars; Freiesleben Hansen, Per

    1999-01-01

    Couplings between creep of hardened concrete and temperature/water effects are well-known. Both the level and the gradients in time of temperature or water content influence the creep properties. In early age concrete the internal drying and the heat development due to hydration increase the effe...... of experimental results for creep of early age and hardened concrete either at different constant temperature levels or for varuing temperature histories illustrate the model....

  6. Tensile creep of beta phase zircaloy-2

    International Nuclear Information System (INIS)

    Burton, B.; Reynolds, G.L.; Barnes, J.P.

    1977-08-01

    The tensile creep and creep rupture properties of beta-phase zircaloy-2 are studied under vacuum in the temperature and stress range 1300-1550 K and 0.5-2 MN/m 2 . The new results are compared with previously reported uniaxial and biaxial data. A small but systematic difference is noted between the uniaxial and biaxial creep data and reasons for this discrepancy are discussed. (author)

  7. Modeling of creep for structural analysis

    Energy Technology Data Exchange (ETDEWEB)

    Naumenko, K.; Altenbach, H. [Halle-Wittenberg Univ., Halle (Germany). Lehrstuhl fuer Technische Mechanik

    2007-07-01

    ''Creep Modeling for Structural Analysis'' develops methods to simulate and analyze the time-dependent changes of stress and strain states in engineering structures up to the critical stage of creep rupture. The principal subjects of creep mechanics are the formulation of constitutive equations for creep in structural materials under multi-axial stress states; the application of structural mechanics models of beams, plates, shells and three-dimensional solids and the utilization of procedures for the solution of non-linear initial-boundary value problems. The objective of this book is to review some of the classical and recently proposed approaches to the modeling of creep for structural analysis applications as well as to extend the collection of available solutions of creep problems by new, more sophisticated examples. In Chapter 1, the book discusses basic features of the creep behavior in materials and structures and presents an overview of various approaches to the modeling of creep. Chapter 2 collects constitutive models that describe creep and damage processes under multi-axial stress states. Chapter 3 deals with the application of constitutive models to the description of creep for several structural materials. Constitutive and evolution equations, response functions and material constants are presented according to recently published experimental data. In Chapter 4 the authors discuss structural mechanics problems. Governing equations of creep in three-dimensional solids, direct variational methods and time step algorithms are reviewed. Examples are presented to illustrate the application of advanced numerical methods to the structural analysis. An emphasis is placed on the development and verification of creep-damage material subroutines inside the general purpose finite element codes. (orig.)

  8. Viscoelastic creep of high-temperature concrete

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Marchertas, A.H.; Bazant, Z.P.

    1985-01-01

    Presented in this report is the analytical model for analysis of high temperature creep response of concrete. The creep law used is linear (viscoelastic), the temperature and moisture effects on the creep rate and also aging are included. Both constant and transient temperature as well as constant and transient moisture conditions are considered. Examples are presented to correlate experimental data with parameters of the analytical model by the use of a finite element scheme

  9. Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

    International Nuclear Information System (INIS)

    Grover, S.B.; Metcalfe, M.P.

    2002-01-01

    A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment

  10. Model for transient creep of southeastern New Mexico rock salt

    International Nuclear Information System (INIS)

    Herrmann, W.; Wawersik, W.R.; Lauson, H.S.

    1980-11-01

    In a previous analysis, existing experimental data pertaining to creep tests on rock salt from the Salado formation of S.E. New Mexico were fitted to an exponential transient creep law. While very early time portions of creep strain histories were not fitted very well for tests at low temperatures and stresses, initial creep rates in particular generally being underestimated, the exponential creep law has the property that the transient creep strain approaches a finite limit with time, and is therefore desirable from a creep modelling point of view. In this report, an analysis of transient creep is made. It is found that exponential transient creep can be related to steady-state creep through a universal creep curve. The resultant description is convenient for creep analyses where very early time behavior is not important

  11. Study on the creep constitutive equation of Hastelloy X, (1)

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Mutoh, Yasushi

    1983-01-01

    A creep constitutive equation of Hastelloy X was obtained from available experimental data. A sensitivity analysis of this creep constitutive equation was carried out. As the result, the following were revealed: (i) Variations in creep behavior with creep constitutive equation are not small. (ii) In a simpler stress change pattern, variations in creep behavior are similar to those in the corresponding fundamental creep characteristics (creep strain curve, stress relaxation curve, etc.). (iii) Cumulative creep damage estimated in accordance with ASME Boiler and Pressure Vessel Code Case N-47 from a stress history predicted by ''the standard creep constitutive equation'' which predicts the average behavior of creep strain curve data is not thought to be on the safe side on account of uncertainties in creep damage caused by variations in creep strain curve. (author)

  12. Modification of PMMA/graphite nanocomposites through ion beam technique

    Science.gov (United States)

    Singhal, Prachi; Rattan, Sunita; Avasthi, Devesh Kumar; Tripathi, Ambuj

    2013-08-01

    Swift heavy ion (SHI) irradiation is a special technique for inducing physical and chemical modifications in bulk materials. In the present work, the SHI hs been used to prepare nanocomposites with homogeneously dispersed nanoparticles. The nanographite was synthesized from graphite using the intercalation-exfoliation method. PMMA Poly(methyl methacrylate)/graphite nanocomposites have been synthesized by in situ polymerization. The prepared PMMA/graphite nanocomposite films were irradiated with SHI irradiation (Ni ion beam, 80 MeV and C ion beam, 50 MeV) at a fluence of 1×1010 to 3×1012 ions/cm2. The nanocomposite films were characterized by scanning electron microscope (SEM) and were evaluated for their electrical and sensor properties. After irradiation, significant changes in surface morphology of nanocomposites were observed as evident from the SEM images, which show the presence of well-distributed nanographite platelets. The irradiated nanocomposites exhibit better electrical and sensor properties for the detection of nitroaromatics with marked improvement in sensitivity as compared with unirradiated nanocomposites.

  13. A German research project about applicable graphite cutting techniques

    International Nuclear Information System (INIS)

    Holland, D.; Quade, U.; Bach, F.W.; Wilk, P.

    2001-01-01

    In Germany, too, quite large quantities of irradiated nuclear graphite, used in research and prototype reactors, are waiting for an environmental way of disposal. While incineration of nuclear graphite does not seem to be a publicly acceptable way, cutting and packaging into ductile cast iron containers could be a suitable way of disposal in Germany. Nevertheless, the cutting of graphite is also a very difficult technique by which a large amount of secondary waste or dust might occur. An applicable graphite cutting technique is needed. Therefore, a group of 13 German partners, consisting of one university, six research reactor operators, one technical inspection authority, three engineering companies, one industrial cutting specialist and one commercial dismantling company, decided in 1999 to start a research project to develop an applicable technique for cutting irradiated nuclear graphite. Aim of the project is to find the most suitable cutting techniques for the existing shapes of graphite blocks with a minimum of waste production rate. At the same time it will be learned how to sample the dust and collect it in a filter system. The following techniques will be tested and evaluated: thermal cutting, water jet cutting, mechanical cutting with a saw, plasma arc cutting, drilling. The subsequent evaluation will concentrate on dust production, possible irradiation of staff, time and practicability under different constraints. This research project is funded by the German Minister of Education and Research under the number 02 S 7849 for a period of two years. A brief overview about the work to be carried out in the project will be given. (author)

  14. Harwell Graphite Calorimeter

    International Nuclear Information System (INIS)

    Linacre, J.K.

    1970-01-01

    The calorimeter is of the steady state temperature difference type. It contains a graphite sample supported axially in a graphite outer jacket, the assembly being contained in a thin stainless steel outer can. The temperature of the jacket and the temperature difference between sample and jacket are measured by chromel-alumel thermocouples. The instrument is calibrated by means of an electric heater of low mass positioned on the axis of the sample. The resistance of the heater is known and both current through the heater and the potential across it may be measured. The instrument is filled with nitrogen at a pressure of one half atmosphere at room temperature. The calorimeter has been designed for prolonged operation at temperatures up to 600°C, and dose rates up to 1 Wg -1 , and instruments have been in use for periods in excess of one year

  15. Thermal migration of deuterium implanted in graphite: Influence of free surface proximity and structure

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Moncoffre, N., E-mail: n.moncoffre@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Ammar, M.R. [CNRS, CEMHTI UPR3079, Université Orléans, CS90055, F-45071 Orléans cedex 2 (France); Rouzaud, J.N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure, Paris, UMR CNRS ENS 8538, F-75231 Paris cedex 5 (France)

    2016-03-15

    This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO{sub 2} cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges R{sub p} of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D({sup 3}He,p){sup 4}He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200–1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower

  16. Thermal migration of deuterium implanted in graphite: Influence of free surface proximity and structure

    Science.gov (United States)

    Le Guillou, M.; Moncoffre, N.; Toulhoat, N.; Pipon, Y.; Ammar, M. R.; Rouzaud, J. N.; Deldicque, D.

    2016-03-01

    This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO2 cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges Rp of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D(3He,p)4He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200-1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower temperatures when D is located

  17. Thermal Creep Force: Analysis And Application

    Science.gov (United States)

    2016-06-01

    The boundary condition was inflow and outflow so particles whose trajectory took them outside the simulation space would no longer be simulated and...Calhoun: The NPS Institutional Archive Theses and Dissertations Thesis and Dissertation Collection 2016-06 Thermal creep force: analysis and...CALIFORNIA DISSERTATION Approved for public release; distribution is unlimited THERMAL CREEP FORCE: ANALYSIS AND APPLICATION by David

  18. Making Ice Creep in the Classroom

    Science.gov (United States)

    Prior, David; Vaughan, Matthew; Banjan, Mathilde; Hamish Bowman, M.; Craw, Lisa; Tooley, Lauren; Wongpan, Pat

    2017-04-01

    Understanding the creep of ice has direct application to the role of ice sheet flow in sea level and climate change and to modelling of icy planets and satellites of the outer solar system. Additionally ice creep can be used as an analogue for the high temperature creep of rocks, most particularly quartzites. We adapted technologies developed for ice creep experiments in the research lab, to build some inexpensive ( EU200) rigs to conduct ice creep experiments in an undergraduate (200 and 300 level) class in rock deformation. The objective was to give the students an experience of laboratory rock deformation experiments so that they would understand better what controls the creep rate of ice and rocks. Students worked in eight groups of 5/6 students. Each group had one deformation rig and temperature control system. Each group conducted two experiments over a 2 week period. The results of all 16 experiments were then shared so that all students could analyse the mechanical data and generate a "flow law" for ice. Additionally thin sections were made of each deformed sample so that some microstructural analysis could be incorporated in the data analysis. Students were able to derive a flow law that showed the relationship of creep rate to both stress and temperature. The flow law matches with those from published research. The class did provide a realistic introduction to laboratory rock deformation experiments and helped students' understanding of what controls the creep of rocks.

  19. Micro-orientation control of silicon polymer thin films on graphite surfaces modified by heteroatom doping

    Energy Technology Data Exchange (ETDEWEB)

    Shimoyama, Iwao, E-mail: shimoyama.iwao@jaea.go.jp [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Baba, Yuji [Fukushima Administrative Department, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Hirao, Norie [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan)

    2017-05-31

    Highlights: • Micro-orientation control method for organic polysilane thin films is proposed. • This method utilizes surface modification of graphite using heteroatom doping. • Lying, standing, and random orientations can be freely controlled by this method. • Micro-pattering of a polysilane film with controlled orientations is achieved. - Abstract: Near-edge X-ray absorption fine structure (NEXAFS) spectroscopy is applied to study orientation structures of polydimethylsilane (PDMS) films deposited on heteroatom-doped graphite substrates prepared by ion beam doping. The Si K-edge NEXAFS spectra of PDMS show opposite trends of polarization dependence for non irradiated and N{sub 2}{sup +}-irradiated substrates, and show no polarization dependence for an Ar{sup +}-irradiated substrate. Based on a theoretical interpretation of the NEXAFS spectra via first-principles calculations, we clarify that PDMS films have lying, standing, and random orientations on the non irradiated, N{sub 2}{sup +}-irradiated, and Ar{sup +}-irradiated substrates, respectively. Furthermore, photoemission electron microscopy indicates that the orientation of a PDMS film can be controlled with microstructures on the order of μm by separating irradiated and non irradiated areas on the graphite surface. These results suggest that surface modification of graphite using ion beam doping is useful for micro-orientation control of organic thin films.

  20. Creep deformation of restorative resin-composites intended for bulk-fill placement.

    Science.gov (United States)

    El-Safty, S; Silikas, N; Watts, D C

    2012-08-01

    To determine the creep deformation of several "bulk-fill" resin-composite formulations in comparison with some other types. Six resin-composites; four bulk-fill and two conventional were investigated. Stainless steel split molds (4 mm × 6 mm) were used to prepare cylindrical specimens for creep testing. Specimens were thoroughly irradiated with 650 mW cm(-2). A total of 10 specimens for each material were divided into two groups (n = 5) according to the storage condition; Group A stored dry at 37 °C for 24h and Group B stored in distilled water at 37 °C in an incubator for 24h. Each specimen was loaded (20 MPa) for 2h and unloaded for 2h. The strain deformation was recorded continuously for 4h. Statistical analysis was performed using a two-way ANOVA followed by one-way ANOVA and the Bonferroni post hoc test at a significance level of a = 0.05. The maximum creep strain % ranged from 0.72% up to 1.55% for Group A and the range for Group B increased from 0.79% up to 1.80% due to water sorption. Also, the permanent set ranged from 0.14% up to 0.47% for Group A and from 0.20% up to 0.59% for Group B. Dependent on the material and storage condition, the percentage of creep strain recovery ranged between 64% and 81%. Increased filler loading in the bulk-fill materials decreased the creep strain magnitude. Creep deformation of all studied resin-composites increased with wet storage. The "bulk-fill" composites exhibited an acceptable creep deformation and within the range exhibited by other resin-composites. Copyright © 2012 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

  1. Creep resistant high temperature martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

    2015-11-13

    The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6 carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followed by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.

  2. Creep resistant high temperature martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

    2017-01-31

    The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, copper, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followed by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.

  3. Statistical analysis of concrete creep effects

    International Nuclear Information System (INIS)

    Floris, C.

    1989-01-01

    The principal sources of uncertainty in concrete creep effects are the following: uncertainty in the stochastic evolution in time of the mechanism of creep (internal uncertainty); uncertainty in the prediction of the properties of the materials; uncertainty in the stochastic evolution of environmental conditions; uncertainty of the theoretical models; errors of measurement. Interest in the random nature of concrete creep (and shrinkage) effects is discussed. The late beginning of the studies on this subject is perhaps due to their theoretical and computational complexity: nevertheless, since creep and shrinkage affect features of concrete structures as the residual prestressing force in prestressed sections, the stress redistribution in steel-concrete composite beams, deflections and deformations, stress distributions in non-homogenous structures, reactions due to delayed restraints and creep buckling, these studies are very important. This paper is aimed to find the statistics of some of these effects taking into the account the third type of source of uncertainty

  4. Assessment of concrete creep and shrinkage

    International Nuclear Information System (INIS)

    Trivedi, Neha; Singh, R.K.

    2012-01-01

    B-3 model prediction of concrete creep and shrinkage strains on cylindrical specimen and BARC Containment test model (BARCOM) are presented. Experimental shrinkage strain is shown to be in agreement with B-3 model predictions for cylindrical specimen and BARCOM. Creep strain in cylindrical specimen is found to be in agreement with B-3 model. In BARCOM for wall cast in different pores, creep strain is in agreement with B-3 model in hoop direction however in longitudinal direction, observed creep strain in higher than B-3 model. For dome structure cast in a single pour, experimental creep strain shows confirmity with B-3 model both in hoop and longitudinal directions. The study on concrete aging and average longitudinal shrinkage strain is carried out. (author)

  5. Creep and relaxation behavior of Inconel-617

    International Nuclear Information System (INIS)

    Osthoff, W.; Ennis, P.J.; Nickel, H.; Schuster, H.

    1984-01-01

    The static and dynamic creep behavior of Inconel alloy 617 has been determined in constant load creep tests, relaxation tests, and stress reduction tests in the temperature range 1023 to 1273 K. The results have been interpreted using the internal stress concept: The dependence of the internal stress on the applied stress and test temperature was determined. In a few experiments, the influence of cold deformation prior to the creep test on the magnitude of the internal stress was also investigated. It was found that the experimentally observed relaxation behavior could be more satisfactorily described using the Norton creep equation modified by incorporation of the internal stress than by the conventional Norton creep equation

  6. Deformation mechanisms in cyclic creep and fatigue

    International Nuclear Information System (INIS)

    Laird, C.

    1979-01-01

    Service conditions in which static and cyclic loading occur in conjunction are numerous. It is argued that an understanding of cyclic creep and cyclic deformation are necessary both for design and for understanding creep-fatigue fracture. Accordingly a brief, and selective, review of cyclic creep and cyclic deformation at both low and high strain amplitudes is provided. Cyclic loading in conjunction with static loading can lead to creep retardation if cyclic hardening occurs, or creep acceleration if softening occurs. Low strain amplitude cyclic deformation is understood in terms of dislocation loop patch and persistent slip band behavior, high strain deformation in terms of dislocation cell-shuttling models. While interesting advances in these fields have been made in the last few years, the deformation mechanisms are generally poorly understood

  7. Structural disorder of graphite and implications for graphite thermometry

    Directory of Open Access Journals (Sweden)

    M. Kirilova

    2018-02-01

    Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  8. GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

    Energy Technology Data Exchange (ETDEWEB)

    Marquez, Eva; Pina, Gabriel; Rodriguez, Marina [CIEMAT, Av. Complutense, 22, 28040-MADRID (Spain); Fachinger, Johannes; Grosse, Karl-Heinz [Furnaces Nuclear Application Grenoble SAS (FNAG), 4, avenue Charles de Gaulle, 38800 Le Pont de Claix (France); Leganes Nieto, Jose Luis; Quiros Gracian, Maria [ENRESA, C/ Emilio Vargas,7 - 28043 - MADRID (Spain); Seemann, Richard [ALD Vacuum Technologies GmbH, Wilhelm-Rohn-Strasse 35, 63450 Hanau (Germany)

    2012-07-01

    Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the

  9. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram

    2017-07-20

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a mechanical pressing operation to generate a bromine-graphite/metal composite material.

  10. Chemical stabilization of graphite surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Bistrika, Alexander A.; Lerner, Michael M.

    2018-04-03

    Embodiments of a device, or a component of a device, including a stabilized graphite surface, methods of stabilizing graphite surfaces, and uses for the devices or components are disclosed. The device or component includes a surface comprising graphite, and a plurality of haloaryl ions and/or haloalkyl ions bound to at least a portion of the graphite. The ions may be perhaloaryl ions and/or perhaloalkyl ions. In certain embodiments, the ions are perfluorobenzenesulfonate anions. Embodiments of the device or component including stabilized graphite surfaces may maintain a steady-state oxidation or reduction surface current density after being exposed to continuous oxidation conditions for a period of at least 1-100 hours. The device or component is prepared by exposing a graphite-containing surface to an acidic aqueous solution of the ions under oxidizing conditions. The device or component can be exposed in situ to the solution.

  11. Impedance of electrochemically modified graphite.

    Science.gov (United States)

    Magdić, Katja; Kvastek, Krešimir; Horvat-Radošević, Višnja

    2014-01-01

    Electrochemical impedance spectroscopy, EIS, has been applied for characterization of electrochemically modified graphite electrodes in the sulphuric acid solution. Graphite modifications were performed by potential cyclization between potentials of graphite oxide formation/reduction, different number of cycles, and prolonged reduction steps after cyclization. Impedance spectra measured at two potential points within double-layer region of graphite have been successfully modeled using the concept of porous electrodes involving two different electrolyte diffusion paths, indicating existence of two classes of pores. The evaluated impedance parameter values show continuous changes with stages of graphite modification, indicating continuous structural changes of pores by number of potential cycles applied. Differences of impedance parameter values at two potential values indicate the potential induced changes of solution properties within the pores of modified graphite.

  12. Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

    2016-04-15

    Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

  13. A FACSIMILE code for calculating void swelling and creep, with vacancy loops present: version VS4

    International Nuclear Information System (INIS)

    Windsor, M.E.; Bullough, R.; Wood, M.H.

    1981-10-01

    This FACSIMILE code calculates void swelling and creep of irradiated materials, taking into account the effects of cavities, interstitial loops, vacancy loops, dislocation network and either grain boundaries or foil surfaces. The creep calculations are based on SIPA theory (stress induced preferred absorption), with no preferred nucleation. Either interactive or non-interactive options are available for the sink strength equations, but rate limitation is not incorporated. FACSIMILE is a computer program for solving simultaneous differential equations, and this VS4 code is one of a series of codes for calculating void swelling using increasingly complex theories. Other reports describing the VS1 and VS2 codes explain their use under control of the TSO system of the Harwell IBM 3033 computer, and explain the basic organization of the codes as required for use by FACSIMILE. The creep theory assumes that the material is under a constant uniaxial tensile stress during the irradiation. Three directions are considered for network parameters relative to the direction of the stress, and two directions for interstitial and vacancy loops. To give a full picture of these various contributions to the total creep, a large set of output parameter values are printed for each demanded dose value via a FORTRAN subroutine. (author)

  14. A model for pulsed laser melting of graphite

    Science.gov (United States)

    Steinbeck, J.; Braunstein, G.; Dresselhaus, M. S.; Venkatesan, T.; Jacobson, D. C.

    1985-12-01

    A model for laser melting of carbon at high temperatures to form liquid carbon has been developed. This model is solved numerically using experimental data from laser irradiation studies in graphite consistent with a melting temperature for graphite of 4300 K. The parameters for high-temperature graphite are based on the extension of previously measured thermal properties into the high-temperature regime. A simple classical free electron gas model is used to calculate the properties of liquid carbon. There is very good agreement between the model calculation and experimental results for laser pulse fluences below 2.0 J/cm2. Modifications to the model for larger laser pulse fluences are discussed.

  15. Heat exchanger using graphite foam

    Science.gov (United States)

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

    A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

  16. Blunt indentation of core graphite

    International Nuclear Information System (INIS)

    Hartley, M.; McEnaney, B.

    1996-01-01

    Blunt indentation experiments were carried out on unoxidized and thermally oxidised IM1-24 graphite as a model to simulate local point stresses acting on graphite moderator bricks. Blunt indentation of unoxidized graphite initiates cracks close to the region of maximum tensile stress at the edge of the indentation. Cracks propagate and converge to form a cone of material. Failure is catastrophic, typically forming three pieces of graphite and ejecting the cone referred to above. The failure mode under indentation loading for highly oxidised graphite (weigh loss > 40%) is different from that for the unoxidized graphite. There is no longer a distinct crack path, the indentation is much deeper than in the case of the unoxidized graphite, and there is a region of crushed debris beneath the indentation, producing a crater-like structure. The reduction in the compressive fracture stress, σ cf , under indentation loading with increasing fractional weight loss on oxidation, x, can be fitted to σ cf /σ 0 = exp-[5.2x] where σ 0 is the compressive fracture stress of the unoxidized graphite. This indicates that the effect of thermal oxidation on indentation fracture stress is more severe than the effects of radiolytic oxidation on conventional strengths of nuclear graphites. (author). 8 refs, 12 figs

  17. Mathematical modelling of the the processes of radiating formation of defects at interaction of carbon with graphite

    Directory of Open Access Journals (Sweden)

    A. Kupchishin

    2012-12-01

    Full Text Available Processes of radiation formation of defects in the carbon irradiated by graphite are considered in work. The regularities arising at selection of approximation expressions, a finding of result area at calculation of cascadely - probabilistic functions depending on number of interactions and depth of penetration of particles are revealed. The regularities formed at calculations of concentration of radiating defects in graphite, irradiated by carbon are received. Results of calculations are presented in the form of tables and schedules.

  18. High dose effects in neutron irradiated face-centered cubic metals

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors

  19. Advances in the assessment of creep data

    Energy Technology Data Exchange (ETDEWEB)

    Holdsworth, S.R.

    2010-07-01

    Many of the classical models representing the creep and rupture behaviour of metals were developed prior to and during the 1950s and 1960s, and their subsequent exploitation, in particular for the assessment of large creep property datasets, was initially limited by the capability of the analytical tools available at the time. The formation of ECCC (the European Creep Collaborative Committee) in 1991, with a main objective of providing reliable peer reviewed long-time creep property values for European Design and Product Standards, led to the development of rigorous assessment procedures such as PD6605 and DESA incorporating post assessment tests to verify: physical realism, effectiveness of model-fit within the range of the source experimental data, and extrapolation credibility. The first ECCC assessment recommendations published in 1996 undoubtedly provided a catalyst for others to exploit the availability of low cost, powerful desktop computers to develop rigorous methodologies for the physically realistic analysis of uniaxial and multi-axial data for the reliable and accurate characterisation of creep strain, and rupture strength and ductility properties. More recent improvements in data assessment methodologies have been driven by the need to effectively model the creep deformation and rupture characteristics of the complex new generation alloys and fabrications being designed to cater for the continually evolving requirements of modern advanced power plant. These advances in the assessment of creep data are reviewed. (orig.)

  20. Creep behavior evaluation of welded joint

    International Nuclear Information System (INIS)

    Susei, Shuzo; Matsui, Shigetomo; Mori, Eisuke; Shimizu, Shigeki; Satoh, Keisuke.

    1980-01-01

    In the creep design of high temperature structural elements, it is necessary to grasp the creep performance of joints as a whole, paying attention to the essential lack of uniformity between the material qualities of parent metals and welds. In this study, the factors controlling the creep performance of butt welded joints were investigated theoretically, when they were subjected to lateral tension and longitudinal tension. It was clarified that the rupture time in the case of laterally pulled joints was determined by the ratio of the creep rupture times of weld metals and parent metals, and the rupture time in the case of longitudinally pulled joints was determined by the ratio of the creep rupture times and the ratio of the creep strain rates of weld metals and parent metals. Moreover, when the joints of the former ratio less than 1 and the latter ratio larger than 1 were investigated experimentally, the rupture time in the case of laterally pulled joints was affected by the relative thickness, and when the relative thickness was large, the theoretical and the experimental values coincided, but the relative thickness was small, the theoretical values gave the evaluation on safe side as compared with the experimental values due to the effect of restricting deformation. In the case of longitudinally pulled joints, the theoretical and the experimental values coincided relatively well. The diagram of classifying the creep performance of welded joints was proposed. (Kako, I.)

  1. A simple model for indentation creep

    Science.gov (United States)

    Ginder, Ryan S.; Nix, William D.; Pharr, George M.

    2018-03-01

    A simple model for indentation creep is developed that allows one to directly convert creep parameters measured in indentation tests to those observed in uniaxial tests through simple closed-form relationships. The model is based on the expansion of a spherical cavity in a power law creeping material modified to account for indentation loading in a manner similar to that developed by Johnson for elastic-plastic indentation (Johnson, 1970). Although only approximate in nature, the simple mathematical form of the new model makes it useful for general estimation purposes or in the development of other deformation models in which a simple closed-form expression for the indentation creep rate is desirable. Comparison to a more rigorous analysis which uses finite element simulation for numerical evaluation shows that the new model predicts uniaxial creep rates within a factor of 2.5, and usually much better than this, for materials creeping with stress exponents in the range 1 ≤ n ≤ 7. The predictive capabilities of the model are evaluated by comparing it to the more rigorous analysis and several sets of experimental data in which both the indentation and uniaxial creep behavior have been measured independently.

  2. Irradiation techniques for carbon materials (Osiris reactor)

    International Nuclear Information System (INIS)

    Genthon, J.P.; Micaud, G.; Mottet, P.

    1976-01-01

    Neutron irradiation devices are described for carbons, graphites and fuel compacts. The use of standard COLIBRI furnaces of different diameters in the OSIRIS core allows long-life irradiations to be carried out, with or without stress, in situ measurements, or fission gas samplings. Examples of the obtained results are given [fr

  3. New irradiation devices at the FRN reactor

    International Nuclear Information System (INIS)

    Stark, W.

    1980-01-01

    In order to fulfill the experimental demands three additional devices were constructed and installed. The first is a vertical irradiation tube in air surrounded by a lead cylinder (in the irradiation position). The second device is a rabbit system ending within the graphite moderator of the thermal column. The third device is so called rotating disk assembly, built to replace the rotary specimen rack

  4. Depleted Hydrocarbon Reservoirs Present a Safe and Practical Burial Solution for Graphite Waste

    International Nuclear Information System (INIS)

    Rahmani, L.

    2016-01-01

    A solution for graphite waste is proposed that combines reliance on thick impermeable host rock that is needed to confine the long-life radioactivity content of most irradiated graphite with low capitalistic and operational unit volume costs that are required to render this bulky waste form manageable. The solution, uniquely applicable to irradiated graphite due to its low dose rates, moderate mechanical strength and light density, consists in three steps: first, graphite is fine-crushed under water; second, it is made in an aqueous suspension; third, the suspension is injected into a deep, disused hydrocarbon reservoir. Each of these steps only involves well mastered techniques. Regulatory changes that may allow this solution to be added to the gamut of available waste routes, geochemical issues, availability of depleted reservoirs and cost projections are presented. (author)

  5. Graphitic packing removal tool

    Science.gov (United States)

    Meyers, Kurt Edward; Kolsun, George J.

    1997-01-01

    Graphitic packing removal tools for removal of the seal rings in one piece. he packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

  6. Low-temperature creep of austenitic stainless steels

    Science.gov (United States)

    Reed, R. P.; Walsh, R. P.

    2017-09-01

    Plastic deformation under constant load (creep) in austenitic stainless steels has been measured at temperatures ranging from 4 K to room temperature. Low-temperature creep data taken from past and unreported austenitic stainless steel studies are analyzed and reviewed. Creep at cryogenic temperatures of common austenitic steels, such as AISI 304, 310 316, and nitrogen-strengthened steels, such as 304HN and 3116LN, are included. Analyses suggests that logarithmic creep (creep strain dependent on the log of test time) best describe austenitic stainless steel behavior in the secondary creep stage and that the slope of creep strain versus log time is dependent on the applied stress/yield strength ratio. The role of cold work, strain-induced martensitic transformations, and stacking fault energy on low-temperature creep behavior is discussed. The engineering significance of creep on cryogenic structures is discussed in terms of the total creep strain under constant load over their operational lifetime at allowable stress levels.

  7. Creep substructure formation in sodium chloride single crystals in the power law and exponential creep regimes

    Science.gov (United States)

    Raj, S. V.; Pharr, G. M.

    1989-01-01

    Creep tests conducted on NaCl single crystals in the temperature range from 373 to 1023 K show that true steady state creep is obtained only above 873 K when the ratio of the applied stress to the shear modulus is less than or equal to 0.0001. Under other stress and temperature conditions, corresponding to both power law and exponential creep, the creep rate decreases monotonically with increasing strain. The transition from power law to exponential creep is shown to be associated with increases in the dislocation density, the cell boundary width, and the aspect ratio of the subgrains along the primary slip planes. The relation between dislocation structure and creep behavior is also assessed.

  8. Compressive creep of silicon nitride

    International Nuclear Information System (INIS)

    Silva, C.R.M. da; Melo, F.C.L. de; Cairo, C.A.; Piorino Neto, F.

    1990-01-01

    Silicon nitride samples were formed by pressureless sintering process, using neodymium oxide and a mixture of neodymium oxide and yttrio oxide as sintering aids. The short term compressive creep behaviour was evaluated over a stress range of 50-300 MPa and temperature range 1200 - 1350 0 C. Post-sintering heat treatments in nitrogen with a stepwise decremental variation of temperature were performed in some samples and microstructural analysis by X-ray diffraction and transmission electron microscopy showed that the secondary crystalline phase which form from the remnant glass are dependent upon composition and percentage of aditives. Stress exponent values near to unity were obtained for materials with low glass content suggesting grain boundary diffusion accommodation processes. Cavitation will thereby become prevalent with increase in stress, temperature and decrease in the degree of crystallization of the grain boundary phase. (author) [pt

  9. Modelling of cladding creep collapse

    International Nuclear Information System (INIS)

    Koundy, V.; Forgeron, T.; Hivroz, J.

    1993-01-01

    The effects of the initial ovality and pressure level on the collapse time of Zircaloy-4 tubing subjected to uniform external pressure were examined experimentally and analytically. Experiments were performed on end closed tubes with two metallurgical states: stress relieved and recrystallized. Numerical simulations were accomplished with a specific computer program based on an analytical approach and the calculated results were compared with the experimental ones. As a comparison, the finite element method is also partially examined in this analysis. Numerical collapse times are in good agreement with regard to experimental results in the case of stress relieved structure. They seem to be too conservative in the case of a recrystallized metallurgical state and the use of the anisotropic option ameliorates numerical results. Sensibility of numerical solutions to the formulation of primary creep laws are presented

  10. Demonstration of creep during filtration

    DEFF Research Database (Denmark)

    Christensen, Morten Lykkegaard; Bugge, Thomas Vistisen; Kirchheiner, Anders Løvenbalk

    that the production of filtrate also depends on the characteristic time for the filter cake solids to deform. This is formulated in the Terzaghi-Voigt model in which a secondary consolidation is introduced. The secondary consolidation may be visualized by plots of the relative cake deformation (U) v.s. the square...... root of time. Even more clearly it is demonstrated by plotting the liquid pressure at the cake piston interface v.s. the relative deformation (to be shown). The phenomenon of a secondary consolidation processes is in short called creep. Provided that the secondary consolidation rate is of the same......The classical filtration theory assumes a unique relationship between the local filter cake porosity and the local effective pressure. For a number of compressible materials, it has however been observed that during the consolidation stage this may not be the case. It has been found...

  11. Heavy-ion irradiation induced diamond formation in carbonaceous materials

    International Nuclear Information System (INIS)

    Daulton, T. L.

    1999-01-01

    The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

  12. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  13. Comparison of low stress creep properties of ferritic and austenitic creep resistant steels

    Czech Academy of Sciences Publication Activity Database

    Kloc, Luboš; Sklenička, Václav; Ventruba, J.

    319-321, - (2001), s. 774-778 ISSN 0921-5093. [International Conference on Strength of Materials /12./. Monterey, CA, USA, 27.08.2000-01.09.2000] R&D Projects: GA AV ČR IAA2041702; GA MŠk OC 522.40 Institutional research plan: CEZ:AV0Z2041904 Keywords : viscous creep * power-law creep * creep-resistant steel Subject RIV: JG - Metallurgy Impact factor: 0.978, year: 2001

  14. Testing of Small Graphite Samples for Nuclear Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Julie Chapman

    2010-11-01

    Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

  15. Damaging by fatigue and creep of PWR fuel cans. Programme and work in progress

    International Nuclear Information System (INIS)

    Brun, G.

    1983-06-01

    The experimental programme consists in the study of rods, tubes and irradiated cans of zircaloy 4. Up to now only rods have been examined. Tensile properties, creep, low cycle fatigue and microstructure of industrial zircaloy 4 are determined at 20 0 C and 350 0 C with recrystallyzed or annealed material. Results are compared with those of litterature but more results are needed for a statistical analysis [fr

  16. Some aspects of nuclear graphite production in France; Etude generale sur les graphites nucleaires produits en France

    Energy Technology Data Exchange (ETDEWEB)

    Gueron, J.; Hering, H. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legendre, A. [Pechiney, 75 - Paris (France)

    1958-07-01

    1) Manufacturing: A summary and results on the CEA-Pechiney purification process are given. Variations in the preparation of green pastes and their effects on graphitized material are described. 2) Physical and mechanical properties: Results are given on: - Statistics of dimensional variatior products having square cross-section. - Statistical variation of thermal expansion coefficients and of electrical conductivity. - Density of normals to carbon layer planes and their connexion with thermal expansion. - Stress-strain cycles and conclusions drawn therefrom. - Mechanical resistance and gas permeability of items for supporting fuel elements. 3) Behaviour under radiation: Alteration under radiation of French graphites irradiated either in G1 pile or in experimental piles, and thermal annealing of those alterations, are given. (author)Fren. [French] 1) Fabrication: On resume le procede d'epuration CEA-PECHINEY, ainsi que diverses modalites de preparation des pates et on expose les resultats obtenus. 2) Proprietes physiques et mecaniques: On indique le resultat d'etudes sur: - la statistique des dimensions de produits a section carree. - celle des variations des coefficients de dilatation thermique et de la conductibilite electrique. - la densite des normales aux plans graphitiques et leur connexion avec la dilatation thermique. - la compression mecanique du graphite. - la solidite mecanique et la permeabilite aux gaz de pieces destinees a supporter des cartouches de combustible. 3) Tenue sous rayonnement: Modification sous rayonnement des graphites fran is irradies soit dans la pile G1, soit dans des piles experimentales, et guerison thermique de ces modifications. (auteur)

  17. Glass-Graphite Composite Materials

    International Nuclear Information System (INIS)

    Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

    2016-01-01

    A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

  18. Investigation of carbon near the graphite-diamond-liquid triple point

    International Nuclear Information System (INIS)

    Prawer, S.; Jamieson, D.N.

    1992-01-01

    Pulsed laser irradiation is used to heat deeply buried damage layers in diamond. Over a small range of laser powers, damage annealing, formation of buried graphitic layers, and melting of diamond followed by its conversion upon cooling into graphite are observed. The diagnostics employed are Channeling Contrast Microscopy, optical absorption, surface profilometry, and scanning and optical microscopies. The results are explained in terms of the behaviour of carbon under high internal pressures close to the diamond-graphite-liquid carbon triple point in the phase diagram. 17 refs., 3 figs

  19. Slow creep in soft granular packings.

    Science.gov (United States)

    Srivastava, Ishan; Fisher, Timothy S

    2017-05-14

    Transient creep mechanisms in soft granular packings are studied numerically using a constant pressure and constant stress simulation method. Rapid compression followed by slow dilation is predicted on the basis of a logarithmic creep phenomenon. Characteristic scales of creep strain and time exhibit a power-law dependence on jamming pressure, and they diverge at the jamming point. Microscopic analysis indicates the existence of a correlation between rheology and nonaffine fluctuations. Localized regions of large strain appear during creep and grow in magnitude and size at short times. At long times, the spatial structure of highly correlated local deformation becomes time-invariant. Finally, a microscale connection between local rheology and local fluctuations is demonstrated in the form of a linear scaling between granular fluidity and nonaffine velocity.

  20. Implications of Jeffreys-Lomnitz Transient Creep

    Science.gov (United States)

    Strick, Ellis

    1984-01-01

    In 1958 Jeffreys proposed a power law generalization of the logarithmic transient creep earlier attributed to Lomnitz. Although Jeffreys' power law form was admittedly defective in that it became unbounded at infinite time, he did apply it to the viscoelastic behavior of the earth-moon system. Since then it has been successfully applied by many investigators to mantle rehology and Chandler wobble. Experimental seismic studies indicate that most rock types exhibit the almost constant Q behavior which Lomnitz showed to be associated with his logarithmic creep. In this paper, we study not only the Q behavior related to Jeffreys' power law creep but also other mechanical properties such as a precise spring-dashpot ladder network realization are developed. In addition, a very simple physically realizable modification of this ladder network leads to a boundedness at long times of Jeffreys' creep in a manner which does not affect his successful application at finite times.

  1. Creep-fatigue of low cobalt superalloys

    Science.gov (United States)

    Halford, G. R.

    1982-01-01

    Testing for the low cycle fatigue and creep fatigue resistance of superalloys containing reduced amounts of cobalt is described. The test matrix employed involves a single high temperature appropriate for each alloy. A single total strain range, again appropriate to each alloy, is used in conducting strain controlled, low cycle, creep fatigue tests. The total strain range is based upon the level of straining that results in about 10,000 cycles to failure in a high frequency (0.5 Hz) continuous strain-cycling fatigue test. No creep is expected to occur in such a test. To bracket the influence of creep on the cyclic strain resistance, strain hold time tests with ore minute hold periods are introduced. One test per composition is conducted with the hold period in tension only, one in compression only, and one in both tension and compression. The test temperatures, alloys, and their cobalt compositions that are under study are given.

  2. Critical view on the creep modelling procedures

    Czech Academy of Sciences Publication Activity Database

    Kloc, Luboš

    2015-01-01

    Roč. 128, č. 4 (2015), s. 540-542 ISSN 0587-4246. [ISPMA 2014 - International Symposium on Physics of Materials /13./. Praha, 31.08.2014-04.09.2014] R&D Projects: GA MPO FR-TI4/406 Institutional support: RVO:68081723 Keywords : Creep * Creep deformation * Grain boundaries * Phase structure * Strain rate Subject RIV: JJ - Other Materials Impact factor: 0.525, year: 2015

  3. Numerically and experimentally analysis of creep

    International Nuclear Information System (INIS)

    Fontanive, J.A.

    1982-11-01

    The problems of creep in concrete are analyzed experimentally and numerically, comparing with classical methods and suggesting a numerical procedure for the solution of these problems. Firstly, fundamentals of viscoelasticity and its application to concrete behaviour representation are presented. Then the theories of Dischinger and Arutyunyan are studied, and a computing numerical solutions are compared in several examples. Finally, experiences on creep and relaxation are described, and its result are analyzed. Some coments on possible future developments are included. (Author) [pt

  4. Surface analysis of graphite fiber reinforced polyimide composites

    Science.gov (United States)

    Messick, D. L.; Progar, D. J.; Wightman, J. P.

    1983-01-01

    Several techniques have been used to establish the effect of different surface pretreatments on graphite-polyimide composites. Composites were prepared from Celion 6000 graphite fibers and the polyimide LARC-160. Pretreatments included mechanical abrasion, chemical etching and light irradiation. Scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS) were used in the analysis. Contact angle of five different liquids of varying surface tensions were measured on the composites. SEM results showed polymer-rich peaks and polymer-poor valleys conforming to the pattern of the release cloth used durng fabrication. Mechanically treated and light irradiated samples showed varying degrees of polymer peak removal, with some degradation down to the graphite fibers. Minimal changes in surface topography were observed on concentrations of surface fluorine even after pretreatment. The light irradiation pretreatment was most effective at reducing surface fluorine concentrations whereas chemical pretreatment was the least effective. Critical surface tensions correlated directly with the surface fluorine to carbon ratios as calculated from XPS.

  5. Creep analysis of silicone for podiatry applications.

    Science.gov (United States)

    Janeiro-Arocas, Julia; Tarrío-Saavedra, Javier; López-Beceiro, Jorge; Naya, Salvador; López-Canosa, Adrián; Heredia-García, Nicolás; Artiaga, Ramón

    2016-10-01

    This work shows an effective methodology to characterize the creep-recovery behavior of silicones before their application in podiatry. The aim is to characterize, model and compare the creep-recovery properties of different types of silicone used in podiatry orthotics. Creep-recovery phenomena of silicones used in podiatry orthotics is characterized by dynamic mechanical analysis (DMA). Silicones provided by Herbitas are compared by observing their viscoelastic properties by Functional Data Analysis (FDA) and nonlinear regression. The relationship between strain and time is modeled by fixed and mixed effects nonlinear regression to compare easily and intuitively podiatry silicones. Functional ANOVA and Kohlrausch-Willians-Watts (KWW) model with fixed and mixed effects allows us to compare different silicones observing the values of fitting parameters and their physical meaning. The differences between silicones are related to the variations of breadth of creep-recovery time distribution and instantaneous deformation-permanent strain. Nevertheless, the mean creep-relaxation time is the same for all the studied silicones. Silicones used in palliative orthoses have higher instantaneous deformation-permanent strain and narrower creep-recovery distribution. The proposed methodology based on DMA, FDA and nonlinear regression is an useful tool to characterize and choose the proper silicone for each podiatry application according to their viscoelastic properties. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  7. Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2011-08-15

    Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the

  8. Study on the creep constitutive equation of Hastelloy X, (1)

    International Nuclear Information System (INIS)

    Suzuki, Kazuhiko; Mutoh, Yasushi

    1983-01-01

    In order to carry out the structural design of high temperature pipings, intermediate heat exchangers and isolating valves for a multipurpose high temperature gas-cooled reactor, in which coolant temperature reaches 1000 deg C, the creep characteristics of Hastelloy X used as the heat resistant material must be clarified. In addition to usual creep rupture life and the time to reach a specified creep strain, the dependence of creep strain curves on time, temperature and stress must be determined and expressed with equations. Therefore, using the creep data of Hastelloy X given in the literatures, the creep constitutive equation was made. Since the creep strain curves under the same test condition were different according to heats, the sensitivity analysis of the creep constitutive equation was performed. The form of the creep constitutive equation was determined to be Garofalo type. The result of the sensitivity analysis is reported. (Kako, I.)

  9. Acoustic emission from polycrystalline graphites

    International Nuclear Information System (INIS)

    Ioka, I.; Yoda, S.; Oku, T.; Miyamoto, Y.

    1987-01-01

    Acoustic emission was monitored from polycrystalline graphites with different microstructure (pore size and pore volume) subjected to compressive loading. The graphites used in this study comprised five brands, that is, PGX, ISEM-1, IG-11, IG-15, and ISO-88. A root mean square (RMS) voltage and event counts of acoustic emission for graphites were measured during compressive loading. The acoustic emission was measured using a computed-based data acquisition and analysis system. The graphites were first deformed up to 80 % of the average fracture stress, then unloaded and reloaded again until the fracture occured. During the first loading, the change in RMS voltage for acoustic emission was detected from the initial stage. During the unloading, the RMS voltage became zero level as soon as the applied stress was released and then gradually rose to a peak and declined. The behavior indicated that the reversed plastic deformation occured in graphites. During the second loading, the RMS voltage gently increased until the applied stress exceeded the maximum stress of the first loading; there is no Kaiser effect in the graphites. A bicrystal model could give a reasonable explanation of this results. The empirical equation between the ratio of σ AE to σ f and σ f was obtained. It is considered that the detection of microfracture by the acoustic emission technique is effective in macrofracture prediction of polycrystalline graphites. (author)

  10. Hypervelocity impacts into graphite

    Science.gov (United States)

    Latunde-Dada, S.; Cheesman, C.; Day, D.; Harrison, W.; Price, S.

    2011-03-01

    Studies have been conducted into the characterisation of the behaviour of commercial graphite (brittle) when subjected to hypervelocity impacts by a range of projectiles. The experiments were conducted with a two-stage gas gun capable of launching projectiles of differing density and strength to speeds of about 6kms-1 at right angles into target plates. The damage caused is quantified by measurements of the crater depth and diameters. From the experimental data collected, scaling laws were derived which correlate the crater dimensions to the velocity and the density of the projectile. It was found that for moderate projectile densities the crater dimensions obey the '2/3 power law' which applies to ductile materials.

  11. The transverse creep deformation and failure characteristics of SCS-6/Ti-6Al-4V metal matrix composites at 482 C

    International Nuclear Information System (INIS)

    Eggleston, M.R.; Ritter, A.M.

    1995-01-01

    While continuous fiber, unidirectional composites are primarily evaluated for their longitudinal properties, the behavior transverse to the fibers often limits their application. In this study, the tensile and creep behaviors of SCS-6/Ti-6Al-4V composites in the transverse direction at 482 C were evaluated. Creep tests were performed in air and argon environments over the stress range of 103 to 276 MPa. The composite was less creep resistant than the matrix when tested at stress values larger than 150 MPa. Below 150 MPa, the composite was ore creep resistant than the unreinforced matrix. Failure of the composite occurred by the ductile propagation of racks emanating from separated fiber interfaces. The environment in which the test was performed affected the creep behavior. At 103 MPa, the creep rate in argon was 4 times slower than the creep rate in air. The SCS-6 silicon-carbide fiber's graphite coating oxidized in the air environment and encouraged the separation of the fiber-matrix interface. However, at high stress levels, the difference in behavior between air- and argon-tested specimens was small. At these stresses, separation of the interface occurred during the initial loading of the composite and the subsequent degradation of the interface did not affect the creep behavior. Finally, the enrichment of the composite's surface by molybdenum during fabrication resulted in an alloyed surface layer that failed in a brittle fashion during specimen elongation. Although this embrittled layer did not appear to degrade the properties of the composite, the existence of a similar layer on a composite with a more brittle matrix might be very detrimental

  12. Radiolytic graphite oxidation revisited

    International Nuclear Information System (INIS)

    Minshall, P.C.; Sadler, I.A.; Wickham, A.J.

    1996-01-01

    The importance of radiolytic oxidation in graphite-moderated CO 2 -cooled reactors has long been recognised, especially in the Advanced Gas-Cooled Reactors where potential rates are higher because of the higher gas pressure and ratings than the earlier Magnox designs. In all such reactors, the rate of oxidation is partly inhibited by the CO produced in the reaction and, in the AGR, further reduced by the deliberate addition of CH 4 . Significant roles are also played by H 2 and H 2 O. This paper reviews briefly the mechanisms of these processes and the data on which they are based. However, operational experience has demonstrated that these basic principles are unsatisfactory in a number of respects. Gilsocarbon graphites produced by different manufacturers have demonstrated a significant difference in oxidation rate despite a similar specification and apparent equivalence in their pore size and distribution, considered to be the dominant influence on oxidation rate for a given coolant-gas composition. Separately, the inhibiting influence of CH 4 , which for many years had been considered to arise from the formation of a sacrificial deposit on the pore walls, cannot adequately be explained by the actual quantities of such deposits found in monitoring samples which frequently contain far less deposited carbon than do samples from Magnox reactors where the only source of such deposits is the CO. The paper also describes the current status of moderator weight-loss predictions for Magnox and AGR Moderators and the validation of the POGO and DIFFUSE6 codes respectively. 2 refs, 5 figs

  13. A creep life assessment method for boiler pipes using small punch creep test

    International Nuclear Information System (INIS)

    Izaki, Toru; Kobayashi, Toshimi; Kusumoto, Junichi; Kanaya, Akihiro

    2009-01-01

    The small punch creep (SPC) test is considered as a highly useful method for creep life assessment for high temperature plant components. SPC uses miniature-sized specimens and does not cause any serious sampling damages, and its assessment accuracy is at a high level. However, in applying the SPC test to the residual creep life assessment of the boiler in service, there are some issues to be studied. In order to apply SPC test to the residual creep life assessment of the 2.25Cr-1Mo steel boiler pipe, the relationship between uniaxial creep stress and the SPC test load has been studied. The virgin material, pre-crept, weldment and service aged samples of 2.25Cr-1Mo steel were tested. It was confirmed that the relationship between uniaxial creep stress and the SPC test load at the same rupture time can be described as a single straight line independent of test conditions and materials. Therefore a life assessment is possible by using SPC test in place of uniaxial creep tests. The creep life assessment using SPC was applied to actual thermal power plant components which are in service.

  14. Graphite structure and magnetic parameters of flake graphite cast iron

    Science.gov (United States)

    Vértesy, G.; Uchimoto, T.; Takagi, T.; Tomáš, I.; Kage, H.

    2017-11-01

    Different matrix and graphite morphologies were generated by a special heat treatment in three chemically different series of flake graphite cast iron samples. As cast, furnace cooled and air cooled samples were investigated. The length of graphite particles and the pearlite volume of samples were determined by metallographic examination and these parameters were compared with the nondestructively measured magnetic parameters. Magnetic measurements were performed by the method of Magnetic Adaptive Testing, which is based on systematic measurement and evaluation of minor magnetic hysteresis loops. It was shown that linear correlation existed between the magnetic quantities and the graphite length, and also between the magnetic quantities and the relative pearlite content in the investigated cast iron. A numerical expression was also determined between magnetic descriptors and relative pearlite content, which does not depend on the detailed experimental conditions.

  15. Eutectic solidification mode of spheroidal graphite cast iron and graphitization

    Directory of Open Access Journals (Sweden)

    Hideo Nakae

    2007-02-01

    Full Text Available The shrinkage and chilling tendency of spheroidal graphite (abbreviated SG cast iron is much greater than that of the flake graphite cast iron in spite of its higher amount of C and Si contents. Why? The main reason should be the difference in their graphitization during the eutectic solidification. In this paper, we discuss the difference in the solidification mechanism of both cast irons for solving these problems using unidirectional solidification and the cooling curves of the spheroidal graphite cast iron. The eutectic solidification rate of the SG cast iron is controlled by the diffusion of carbon through the austenite shell, and the final thickness is 1.4 times the radius of the SG, therefore, the reduction of the SG size, namely, the increase in the number, is the main solution of these problems.

  16. Obtention of nuclear grade graphite

    International Nuclear Information System (INIS)

    Ferreira, M.L.

    1984-01-01

    The impurity level of natural graphite found in some of the most important mines of the State of Minas Gerais - Brasil is determined. It is also concerned with the development and use of natural graphite in nuclear reactors. Standard methods for chemical and instrumentsal analysis such as Spectrografic Determination by Emission, Spectrografic Determination by X-Rays, Spectrografic Determination by Atomic Asorption, Photometric Determination, and also chemical and physical methods for separation of impurities as well standard method for Estimating the Thermal Neutron Absorption Cross Section of graphite were employed. Some aditionals methods of purification to the ordinary treatment such as the use of metanol and halogens are also described. (Author) [pt

  17. AGC-1 Experiment and Final Preliminary Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Robert L. Bratton; Tim Burchell

    2006-08-01

    This report details the experimental plan and design as of the preliminary design review for the Advanced Test Reactor Graphite Creep-1 graphite compressive creep capsule. The capsule will contain five graphite grades that will be irradiated in the Advanced Test Reactor at the Idaho National Laboratory to determine the irradiation induced creep constants. Seven other grades of graphite will be irradiated to determine irradiated physical properties. The capsule will have an irradiation temperature of 900 C and a peak irradiation dose of 5.8 x 10{sup 21} n/cm{sup 2} [E > 0.1 MeV], or 4.2 displacements per atom.

  18. A systematic methodology for creep master curve construction using the stepped isostress method (SSM): a numerical assessment

    Science.gov (United States)

    Miranda Guedes, Rui

    2018-02-01

    Long-term creep of viscoelastic materials is experimentally inferred through accelerating techniques based on the time-temperature superposition principle (TTSP) or on the time-stress superposition principle (TSSP). According to these principles, a given property measured for short times at a higher temperature or higher stress level remains the same as that obtained for longer times at a lower temperature or lower stress level, except that the curves are shifted parallel to the horizontal axis, matching a master curve. These procedures enable the construction of creep master curves with short-term experimental tests. The Stepped Isostress Method (SSM) is an evolution of the classical TSSP method. Higher reduction of the required number of test specimens to obtain the master curve is achieved by the SSM technique, since only one specimen is necessary. The classical approach, using creep tests, demands at least one specimen per each stress level to produce a set of creep curves upon which TSSP is applied to obtain the master curve. This work proposes an analytical method to process the SSM raw data. The method is validated using numerical simulations to reproduce the SSM tests based on two different viscoelastic models. One model represents the viscoelastic behavior of a graphite/epoxy laminate and the other represents an adhesive based on epoxy resin.

  19. Study of graphitic microstructure formation in diamond bulk by pulsed Bessel beam laser writing

    Science.gov (United States)

    Kumar, S.; Sotillo, B.; Chiappini, A.; Ramponi, R.; Di Trapani, P.; Eaton, S. M.; Jedrkiewicz, O.

    2017-11-01

    The advantages of using Bessel beams for the generation of graphitic structures in diamond bulk are presented. We show that by irradiating the sample with a pulsed Bessel beam whose non-diffracting zone is of the same order of the sample thickness, it is possible to produce without any sample translation straight graphitic through-microstructures, whose size depends on the input pulse energy. The microstructure growth is investigated as a function of the number of irradiating pulses, and the femtosecond and picosecond regimes are contrasted.

  20. The investigation of expanded polystyrene creep behaviour

    Directory of Open Access Journals (Sweden)

    Zhukov Aleksey

    2017-01-01

    Full Text Available The results obtained in long-term testing under constant compressive stress of the cut from the Slabs EPS 50/100 and EPS 150 with the density ranging from 15 to 24 kg/m3, which were manufactured by the same manufacturer by foaming EPS solid granules (beads in closed volume. The creep strain of the above described specimens was used as a criterion for estimating the deformability of the EPS slabs under long-term compressive stress. It was measured using special stands EN 1606, maintaining constant stress during the fixed time interval tn=122 days. Creep strains were determined by the methods described in EN 1606 for constant stress σc=0.35σ10% (compressive stress σ10% was determined in accordance with EN 826:2013. The long-term compressive stress measurement error did not exceed 1 %, while the creep strain measurement error was not larger than 0,005 mm. The tests were conducted at the ambient temperature of (23±2°С and relative humidity of (50±5 %.The long-term constant compressive load σc=0.35σ10%. The method of mathematical and statistical experimental design optimization models taking into account the thickness of specimens is proposed to determine the creep compliance Ic (tn the creep strain εc (tn and predictive point estimate of creep strain εc (T. Graphical interpretation of the abstained models is also presented. It should be noted that the abstained equations may be used in practice for estimating the creep strains at time tn=122 days and predictive estimates of εc (T for the load time of 10 years.

  1. Mechanical behaviour of cyclic olefin copolymer/exfoliated graphite nanoplatelets nanocomposites foamed through supercritical carbon dioxide

    Directory of Open Access Journals (Sweden)

    A. Biani

    2016-12-01

    Full Text Available A cycloolefin copolymer matrix was melt mixed with exfoliated graphite nanoplatelets (xGnP and the resulting nanocomposites were foamed by supercritical carbon dioxide. The density of the obtained foams decreased with the foaming pressure. Moreover, xGnP limited the cell growth during the expansion process thus reducing the cell diameter (from 1.08 to 0.22 mm with an XGnP amount of 10 wt% at 150 bar and increasing the cell density (from 12 to 45 cells/mm2 with a nanofiller content of 10 wt% at 150 bar. Electron microscopy observations of foams evidenced exfoliation and orientation of the nanoplatelets along the cell walls. Quasi-static compressive tests and tensile creep tests on foams clearly indicated that xGnP improved the modulus (up to a factor of 10 for a xGnP content of 10 wt% and the creep stability.

  2. Developments in natural uranium - graphite reactors

    International Nuclear Information System (INIS)

    Bourgeois, J.

    1964-01-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  3. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    Science.gov (United States)

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  4. Graphite oxidation in HTGR atmosphere

    International Nuclear Information System (INIS)

    Growcock, F.B.; Barry, J.J.; Finfrock, C.C.; Rivera, E.; Heiser, J.H. III

    1982-01-01

    On-going and recently completed studies of the effect of thermal oxidation on the structural integrity of HTGR candidate graphites are described, and some results are presented and discussed. This work includes the study of graphite properties which may play decisive roles in the graphites' resistance to oxidation and fracture: pore size distribution, specific surface area and impurity distribution. Studies of strength loss mechanisms in addition to normal oxidation are described. Emphasis is placed on investigations of the gas permeability of HTGR graphites and the surface burnoff phenomenon observed during recent density profile measurements. The recently completed studies of catalytic pitting and the effects of prestress and stress on reactivity and ultimate strength are also discussed

  5. Graphite materials for nuclear reactors

    International Nuclear Information System (INIS)

    Oku, Tatsuo

    1991-01-01

    Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

  6. Creep behavior of an epoxy resin and an epoxy-based FRP in condition of simultaneous supply of radiation and stress at cryogenic temperatures

    International Nuclear Information System (INIS)

    Nishiura, Tetsuya; Nishijima, Shigehiro; Okada, Toichi

    1995-01-01

    Creep tests of an epoxy resin and an epoxy-based FRP in bending under irradiation condition have been carried out, to investigate the synergistic effects of radiation and stress on mechanical properties of FRP. Simultaneous supply of stress and irradiation on the epoxy resin and the FRP enhanced creep rates in comparison with that supply of the stress on a post-irradiated one did. ESR spectra measurement was also carried out to study the change of molecule of the resin irradiated. Increase of molecular weight between crosslinks was found out to be enhanced by the synergistic effect of radiation and stress. The mechanism of increased damage of FRP induced by the effects of simultaneous stress and irradiation is discussed. (author)

  7. Graphite oral tattoo: case report

    OpenAIRE

    Moraes, Renata Mendonça; Gouvêa Lima, Gabriela de Morais; Guilhermino, Marinaldo; Vieira, Mayana Soares; Carvalho, Yasmin Rodarte; Anbinder, Ana Lia

    2015-01-01

    Pigmented oral lesions compose a large number of pathological entities, including exogenous pigmentat oral tattoos, such as amalgam and graphite tattoos. We report a rare case of a graphite tattoo on the palate of a 62-year-old patient with a history of pencil injury, compare it with amalgam tattoos, and determine the prevalence of oral tattoos in our Oral Pathology Service. We also compare the clinical and histological findings of grafite and amalgam tattoos. Oral tattoos affect women more f...

  8. Unified creep-plasticity model for halite

    International Nuclear Information System (INIS)

    Krieg, R.D.

    1980-11-01

    There are two national energy programs which are considering caverns in geological salt (NaCl) as a storage repository. One is the disposal of nuclear wastes and the other is the storage of oil. Both short-time and long-time structural deformations and stresses must be predictable for these applications. At 300K, the nominal initial temperature for both applications, the salt is at 0.28 of the melting temperature and exhibits a significant time dependent behavior. A constitutive model has been developed which describes the behavior observed in an extensive set of triaxial creep tests. Analysis of these tests showed that a single deformation mechanism seems to be operative over the stress and temperature range of interest so that the secondary creep data can be represented by a power of the stress over the entire test range. This simple behavior allowed a new unified creep-plasticity model to be applied with some confidence. The resulting model recognizes no inherent difference between plastic and creep strains yet models the total inelastic strain reasonably well including primary and secondary creep and reverse loadings. A multiaxial formulation is applied with a back stress. A Bauschinger effect is exhibited as a consequence and is present regardless of the time scale over which the loading is applied. The model would be interpreted as kinematic hardening in the sense of classical plasticity. Comparisons are made between test data and model behavior

  9. Analysis of Superheater Work Under Creep Conditions

    Directory of Open Access Journals (Sweden)

    Piotr Duda

    2015-03-01

    Full Text Available The aim of this article is work modelling of superheater SH3. It is made of the austenitic stainless steel Super 304H. Its design temperature T is 604 C, and the design pressure P acting on the inner surface of the pipes is 284 bar. The high temperature is the reason of the superheater work under creep conditions. In this article calculations of the optimally mounted coil superheater SH3 are presented. The calculations are carried out first on the basis of the applicable European standards and with the help of the Auto Pipe program. Then, calculations are performed using the ANSYS program based on conducted creep tests and proposed creep equation. The coefficients in creep equation are determined based on the research conducted at the Instytut Metalurgii Żelaza in Gliwice. The model approximates the creep strain as the function of time and stress and this function is presented in the form of a three-dimensional surface . The results of calculations by both methods will be compared and conclusions will be presented. The performed analyzes can estimate the superheater coil remnant life and the usage after the selected time of its operation.

  10. Non Newtonian gravity creeping flow

    International Nuclear Information System (INIS)

    Gratton, J.; Mahajan, S.M.; Minotti, F.

    1988-11-01

    We derive the governing equations for creeping gravity currents of non Newtonian liquids having a power law rheology, using a lubrication approximation. We consider unidirectional and axisymmetric currents. The equations differ from those for Newtonian liquids, being nonlinear in the spatial derivative of the thickness of the current. However, many solutions are closely analogous to those for Newtonian rheology; in particular the spreading relations can also be expressed as power laws of time, with exponents that depend on the rheological index. Similarity solutions for currents whose volume varies as a power of time are obtained. For the spread of a constant volume of liquid, analytic solutions are found. We also derive solutions of the waiting-time type, as well as the ones describing steady flows from a constant source to a sink. General travelling wave solutions are given, and analytic formulae for a simple case are derived. A phase plane formalism, that allows the systematic derivation of self similar solutions, is introduced. The application of the Boltzmann transform is briefly discussed. Present results are closely analogous to those for Newtonian liquids; all the solutions obtained here have their counterparts in Newtonian flows. This happens because the power law rheology, like the Newtonian constitutive relation, involves a single dimensional parameter. Thus one finds similarity solutions whenever the analogous Newtonian problem is self similar. Although the spreading relations are rheology-dependent, in most cases the dependence is rather weak. The present results may be of interest for geophysics since the lithosphere deforms according to an average power law rheology. (author). 17 refs

  11. Creep Properties of Walikukun (Schouthenia ovata Timber Beams

    Directory of Open Access Journals (Sweden)

    Ali Awaludin

    2016-09-01

    Full Text Available This study presents an evaluation of creep constants of Walikukun (Schoutheniaovata timber beams when rheological model of four solid elements, which is obtained byassembling Kelvin and Maxwell bodies in parallel configuration, was adopted. Creep behaviorobtained by this method was further discussed and compared with creep behavior developedusing phenomenological model of the previous study. Creep data of previous study was deformationmeasurement of Walikukun beams having cross-section of 15 mm by 20 mm with a clearspan of 550 mm loaded for three weeks period under two different room conditions: with andwithout Air Conditioner. Creep behavior given by both four solid elements model and phenomenological(in this case are power functions had good agreement during the period of creepmeasurement, but they give different prediction of creep factor beyond this period. The powerfunction of phenomenological model could give a reasonable creep prediction, while for the foursolid elements model a necessary modification is required to adjust its long-term creep behavior.

  12. Viscoelastic characterization of carbon fiber-epoxy composites by creep and creep rupture tests

    International Nuclear Information System (INIS)

    Farina, Luis Claudio

    2009-01-01

    One of the main requirements for the use of fiber-reinforced polymer matrix composites in structural applications is the evaluation of their behavior during service life. The warranties of the integrity of these structural components demand a study of the time dependent behavior of these materials due to viscoelastic response of the polymeric matrix and of the countless possibilities of design configurations. In the present study, creep and creep rupture test in stress were performed in specimens of unidirectional carbon fiber-reinforced epoxy composites with fibers orientations of 60 degree and 90 degree, at temperatures of 25 and 70 degree C. The aim is the viscoelastic characterization of the material through the creep curves to some levels of constant tension during periods of 1000 h, the attainment of the creep rupture envelope by the creep rupture curves and the determination of the transition of the linear for non-linear behavior through isochronous curves. In addition, comparisons of creep compliance curves with a viscoelastic behavior prediction model based on Schapery equation were also performed. For the test, a modification was verified in the behavior of the material, regarding the resistance, stiffness and deformation, demonstrating that these properties were affected for the time and tension level, especially in work temperature above the ambient. The prediction model was capable to represent the creep behavior, however the determination of the equations terms should be considered, besides the variation of these with the applied tension and the elapsed time of test. (author)

  13. Creep Rupture Life Prediction Based on Analysis of Large Creep Deformation

    Directory of Open Access Journals (Sweden)

    YE Wenming

    2016-08-01

    Full Text Available A creep rupture life prediction method for high temperature component was proposed. The method was based on a true stress-strain elastoplastic creep constitutive model and the large deformation finite element analysis method. This method firstly used the high-temperature tensile stress-strain curve expressed by true stress and strain and the creep curve to build materials' elastoplastic and creep constitutive model respectively, then used the large deformation finite element method to calculate the deformation response of high temperature component under a given load curve, finally the creep rupture life was determined according to the change trend of the responsive curve.The method was verified by durable test of TC11 titanium alloy notched specimens under 500 ℃, and was compared with the three creep rupture life prediction methods based on the small deformation analysis. Results show that the proposed method can accurately predict the high temperature creep response and long-term life of TC11 notched specimens, and the accuracy is better than that of the methods based on the average effective stress of notch ligament, the bone point stress and the fracture strain of the key point, which are all based on small deformation finite element analysis.

  14. Modelling fracture of aged graphite bricks under radiation and temperature

    Directory of Open Access Journals (Sweden)

    Atheer Hashim

    2017-05-01

    Full Text Available The graphite bricks of the UK carbon dioxide gas cooled nuclear reactors are subjected to neutron irradiation and radiolytic oxidation during operation which will affect thermal and mechanical material properties and may lead to structural failure. In this paper, an empirical equation is obtained and used to represent the reduction in the thermal conductivity as a result of temperature and neutron dose. A 2D finite element thermal analysis was carried out using Abaqus to obtain temperature distribution across the graphite brick. Although thermal conductivity could be reduced by up to 75% under certain conditions of dose and temperature, analysis has shown that it has no significant effect on the temperature distribution. It was found that the temperature distribution within the graphite brick is non-radial, different from the steady state temperature distribution used in the previous studies [1,2]. To investigate the significance of this non-radial temperature distribution on the failure of graphite bricks, a subsequent mechanical analysis was also carried out with the nodal temperature information obtained from the thermal analysis. To predict the formation of cracks within the brick and the subsequent propagation, a linear traction–separation cohesive model in conjunction with the extended finite element method (XFEM is used. Compared to the analysis with steady state radial temperature distribution, the crack initiation time for the model with non-radial temperature distribution is delayed by almost one year in service, and the maximum crack length is also shorter by around 20%.

  15. Oxidation Protective SiC Coating on Graphite for VHTR Core Support Structure

    International Nuclear Information System (INIS)

    Park, Jae-Won; Kim, Eung-Seon; Kim, Jae-Un; Windes, William E.

    2014-01-01

    The potential for reducing oxidation of the supporting graphite components during normal operation and accident conditions in the VHTR design has been studied. SiC coating on graphite has been studied taking into consideration of possible dimensional change of graphite by the neutron-irradiation. Functionally gradient (FG) SiC coating on the graphite has been performed to moderate the SiC/Graphite interface: E-beam evaporative coating from varied compositions of graphite/SiC mixture in the source crucibles was carried out with an ion beam mixing. The cylindrical graphite samples were uniformly coated by rotating and revolving the samples. Auger depth profile reveals that the ion beam mixed interface is broadened and a cross sectional EDS Si elemental mapping shows a smoothly graded Si profile. The grown film exhibited a stacked columnar structure owing to a frequent sample position change during the coating process, as observed by FE-SEM. As a result of 18 thermal cycling test of 500-1000℃, no film delamination was found on the coated layer, but film cracks were formed, suggesting a strong bonding. When samples were heated at 600°C in static air for 2 h, ~45 wt% of the graphite was burnt off, whereas for the SiC coated graphite only 5 wt %. When heated at 1000 °C in air, vigorous oxidation of graphite took place through a few paths (maybe the mars and/or the crack lines) in the film only leaving the coating layer. As the crack lines were covered with SiC by repeating the ion beam mixed coating process, the oxidation resistance was improved. (author)

  16. Change of Mechanical Properties during Creep Deformation in Modified 9Cr-1Mo Steel

    International Nuclear Information System (INIS)

    Kim, Sung Ho; Han, Chang Hee; Ryu, Woo Seog

    2005-01-01

    9-12% Cr-Mo ferritic/martensitic steels are widely used as high temperature materials in the power plants and chemical industries due to their high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. Owing to the better irradiation characteristics (e.g. excellent irradiation swelling resistance) of these steels than austenitic alloys they have been receiving attention for the application to the fuel cladding or core structure of various advanced nuclear reactors. Since the operating temperature and pressure of advanced nuclear reactors are supposed to be higher than those of light water reactors, high temperature mechanical properties and microstructural stability of cladding and core structural materials of advanced nuclear reactors is very important. Material softening is the main physical phenomenon observed in the crept material. The decrease of the matrix strength by the material softening occurred during creep deformation. When the strength of the matrix decreased to a certain value by creep deformation, the specimen ruptured. The strength changed with applied stress. The ratio of matrix yield strength to applied stress changed from 1.35 at high applied stress to 2.45 at low applied stress. In the present work, we evaluated material softening kinetics by measuring the change of mechanical properties during creep deformation with Indentation typed Tensile Test System (AIS 2000) and Vickers hardness test

  17. High temperature creep strength of Advanced Radiation Resistant Oxide Dispersion Strengthened Steels

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Kim, Tae Kyu

    2014-01-01

    Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, high temperature strength of advanced radiation resistance ODS steel was investigated for the core structural material of next generation nuclear systems. ODS martensitic steel was designed to have high homogeneity, productivity and reproducibility. Mechanical alloying, hot isostactic pressing and hot rolling processes were employed to fabricate the ODS steels, and creep rupture test as well as tensile test were examined to investigate the behavior at high temperatures. ODS steels were fabricated by a mechanical alloying and hot consolidation processes. Mechanical properties at high temperatures were investigated. The creep resistance of advanced radiation resistant ODS steels was more superior than those of ferritic/ martensitic steel, austenitic stainless steel and even a conventional ODS steel

  18. Non-activated high surface area expanded graphite oxide for supercapacitors

    Energy Technology Data Exchange (ETDEWEB)

    Vermisoglou, E.C.; Giannakopoulou, T.; Romanos, G.E.; Boukos, N.; Giannouri, M. [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece); Lei, C.; Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford GU2 7XH (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece)

    2015-12-15

    Graphical abstract: - Highlights: • One-step exfoliation and reduction of graphite oxide via microwave irradiation. • Effect of pristine graphite (type, flake size) on the microwave expanded material. • Effect of pretreatment and oxidation cycles on the produced expanded material. • Expanded graphene materials with high BET surface areas (940 m{sup 2}/g–2490 m{sup 2}/g). • Non-activated graphene based materials suitable for supercapacitors. - Abstract: Microwave irradiation of graphite oxide constitutes a facile route toward production of reduced graphene oxide, since during this treatment both exfoliation and reduction of graphite oxide occurs. In this work, the effect of pristine graphite (type, size of flakes), pretreatment and oxidation cycles on the finally produced expanded material was examined. All the types of graphite that were tested afforded materials with high BET surface areas ranging from 940 m{sup 2}/g to 2490 m{sup 2}/g, without intervening an activation stage at elevated temperature. SEM and TEM images displayed exfoliated structures, where the flakes were significantly detached and curved. The quality of the reduced graphene oxide sheets was evidenced both by X-ray photoelectron spectroscopy and Raman spectroscopy. The electrode material capacitance was determined via electrochemical impedance spectroscopy and cyclic voltammetry. The materials with PEDOT binder had better performance (∼97 F/g) at low operation rates while those with PVDF binder performed better (∼20 F/g) at higher rates, opening up perspectives for their application in supercapacitors.

  19. Transitional Thermal Creep of Early Age Concrete

    DEFF Research Database (Denmark)

    Hauggaard, A. B.; Damkilde, L.; Hansen, Per Freiesleben

    1999-01-01

    Couplings between creep of hardened concrete and temperature/water effects are well-known. Both the level and the gradients in time of temperature or water content influence the creep properties. In early age concrete the internal drying and the heat development due to hydration increase the effect...... of these couplings. The purpose of this work is to set up a mathematical model for creep of concrete that includes the transitional thermal effect. The model governs both early age concrete and hardened concrete. The development of the material properties in the model is assumed to depend on the hydration process...... and the thermal activation of water in the microstructure. The thermal activation is assumed to be governed by the Arrhenius principle, and the activation energy of the viscosity of water is found applicable in the analysis of the experimental data. Changes in temperature create an imbalance in the microstructure...

  20. Documentation for the viscoplastic and creep program

    DEFF Research Database (Denmark)

    Bellini, Anna

    2004-01-01

    The purpose of this document is to summarize the work done in the workpackage 4 of the IDEAL (Integrated Development Routes for Optimized Cast Aluminium Components) project, financed by the EU in frame work 6 and born in collaboration with the automobile and foundry industries. The objective...... of this workpackage is to simulate creep behavior of aluminum cast samples subjected to high temperature. In this document a two-state variables unified model is applied in order to simulate creep behavior and time-dependent metallurgical changes. The fundamental assumption of the unified theory is that creep...... and viscoplasticity, which are both irreversible strains developed because of dislocations motion in the material structure, can be modelled through the implementation of a similar plastic strain velocity law, generally called flow rule. The document shows how to obtain the material data needed for the simulation...

  1. Transitional Thermal Creep of Early Age Concrete

    DEFF Research Database (Denmark)

    Hauggaard-Nielsen, Anders Boe; Damkilde, Lars; Freiesleben Hansen, Per

    1999-01-01

    of these couplings. The purpose of this work is to set up a mathematical model for creep of concrete which includes the transitional thermal effect. The model govern both early age concrete and hardened concrete. The development of the material properties in the model are assumed to depend on the hydration process......Couplings between creep of hardened concrete and temperature/water effects are well-known. Both the level and the gradients in time of temperature or water content influence the creep properties. In early age concrete the internal drying and the heat development due to hydration increase the effect...... and the thermal activation of the water in the microstructure. The thermal activation is assumed to be governed by the Arrhenius principle and the activation energy of the viscosity of water is found applicable in the analysis of experimental data. Changes in temperature create an imbalance in the microstructure...

  2. Influence of the phase composition of refractory materials on creep

    OpenAIRE

    Terzić A.; Pavlović Lj.; Milutinović-Nikolić A.

    2006-01-01

    In this paper, the relationship between the creeping effect and mineralogical characteristics of the applied binding phase for various refractory materials (high-alumina materials, with high or low impurity content, tar bonded either magnesite or dolomite materials and silicate bonded chrom-magnesite materials) is presented. The mechanism of creeping is analyzed and the activation energy for creep for each investigated material is obtained and discussed. All investigated materials are creep s...

  3. Accelerated diffusion controlled creep of polycrystalline materials. Communication 1. Model of diffusion controlled creep acceleration

    International Nuclear Information System (INIS)

    Smirnova, E.S.; Chuvil'deev, V.N.

    1998-01-01

    The model is suggested which describes the influence of large-angle grain boundary migration on a diffusion controlled creep rate in polycrystalline materials (Coble creep). The model is based on the concept about changing the value of migrating boundary free volume when introducing dislocations distributed over the grain bulk into this boundary. Expressions are obtained to calculate the grain boundary diffusion coefficient under conditions of boundary migration and the parameter, which characterized the value of Coble creep acceleration. A comparison is made between calculated and experimental data for Cd, Co and Fe

  4. The effect of creep damage formulation on crack tip fields, creep stress intensity factor and crack growth assessments

    Directory of Open Access Journals (Sweden)

    V. Shlyannikov

    2017-07-01

    Full Text Available Fields of stress, strain rate and process zone of a mode I creep crack growth are analyzed by employing damage evolution equations. Damage models for fracture of process zone represented by stress based formulation. Two expressions are presented to describe the stress-sensitive nature of multiaxial rupture behavior. Both damage free and defective creeping solids have been studied. The variation of creep stress and the crack-tip governing parameter in the form of creep In-integral with time and the evolution of creep damage were analyzed by using the FE-model. The effect of the introduced creep stress intensity factor as a function of creep time through the continuum damage mechanics of the creep crack growth are discussed in detail.

  5. A stochastic approach to anelastic creep

    International Nuclear Information System (INIS)

    Venkataraman, G.

    1976-01-01

    Anelastic creep or the time-dependent yielding or a material subjected to external stresses has been found to be of great importantance in technology in the recent years, particularly in engineering structures including nuclear reactors wherein structural members may be under stress. The physics aspects underlying this phenomenon is dealt with in detail. The basics of time-dependent elasticity, constitutive relation, network models, constitutive equation in the frequency domain and its mearurements, and stochastic approach to creep are discussed. (K.B.)

  6. Radiometric study of creep in ingot rolling

    International Nuclear Information System (INIS)

    Kubicek, P.; Zamyslovsky, Z.; Uherek, J.

    The radiometric study of creep during ingot rolling performed in the rolling mill of the Vitkovice Iron and Steel Works and the first results are described. Selected sites in 3 to 8 ton ingots were labelled with 2 to 3.7x10 5 Bq of 60 Co and after rolling into blocks, the transposition of the labelled sites of the ingots was investigated. The results indicate creep during rolling, local extension in certain sites under study and help to determine the inevitable bottom crop incurred in the forming. Finally, the requirements put on the radiometric apparatus for the next stages of technological research are presented. (author)

  7. Influence of relative humidity on tensile and compressive creep of ...

    African Journals Online (AJOL)

    This paper presents an experimental study on the influence of ambient relative humidity on tensile creep of plain concrete amended with Ground Granulated Blast - furnace Slag and compares it with its influence on compressive creep. Tensile and compressive creep tests were carried out on concrete specimens of 34.49 ...

  8. influence of relative humidity on tensile and compressive creep

    African Journals Online (AJOL)

    HOD

    This paper presents an experimental study on the influence of ambient relative humidity on tensile creep of plain concrete amended with Ground Granulated Blast-furnace Slag and compares it with its influence on compressive creep. Tensile and compressive creep tests were carried out on concrete specimens of ...

  9. Crack Tip Creep Deformation Behavior in Transversely Isotropic Materials

    International Nuclear Information System (INIS)

    Ma, Young Wha; Yoon, Kee Bong

    2009-01-01

    Theoretical mechanics analysis and finite element simulation were performed to investigate creep deformation behavior at the crack tip of transversely isotropic materials under small scale creep (SCC) conditions. Mechanical behavior of material was assumed as an elastic-2 nd creep, which elastic modulus ( E ), Poisson's ratio (v ) and creep stress exponent ( n ) were isotropic and creep coefficient was only transversely isotropic. Based on the mechanics analysis for material behavior, a constitutive equation for transversely isotropic creep behavior was formulated and an equivalent creep coefficient was proposed under plain strain conditions. Creep deformation behavior at the crack tip was investigated through the finite element analysis. The results of the finite element analysis showed that creep deformation in transversely isotropic materials is dominant at the rear of the crack-tip. This result was more obvious when a load was applied to principal axis of anisotropy. Based on the results of the mechanics analysis and the finite element simulation, a corrected estimation scheme of the creep zone size was proposed in order to evaluate the creep deformation behavior at the crack tip of transversely isotropic creeping materials

  10. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  11. Production response of lambs receiving creep feed while grazing ...

    African Journals Online (AJOL)

    Department of Agriculture (Western Cape)

    analysis of variance with treatment (creep feed or no creep feed) and birth status (single and twins) as main factors. Provision a ... At both locations, birth status had no effect on the production parameters for ewes or lambs. Keywords: Creep ... supplementation are of vital importance (De Villiers, 1991; Brand et al., 1999).

  12. Studies of Grain Boundaries in Materials Subjected to Diffusional Creep

    DEFF Research Database (Denmark)

    Nørbygaard, Thomas

    Grain boundaries in crystalline Cu(2%Ni) creep specimens have been studied by use of scanning and transmission electron microscopy in order to establish the mechanism of deformation. Creep rate measurements and dependencies were found to fit reasonably well with the model for diffusional creep...

  13. Plastic creep flow processes in fracture at elevated temperatures

    International Nuclear Information System (INIS)

    Rice, J.R.

    1979-01-01

    Recent theoretical developments on fracture at elevated temperature in the presence of overall plastic (dislocation) creep are discussed. Two topics are considered: stress fields at tips of macroscopic cracks in creeping solids; and diffusive growth of microscopic grain boundary cavities in creeping solids

  14. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  15. In situ and ex situ characterization of the ion-irradiation effects in third generation SiC fibers

    International Nuclear Information System (INIS)

    Huguet-Garcia, Juan

    2015-01-01

    The use of third generation SiC fibers, Tyranno SA3 (TSA3) and Hi Nicalon S (HNS), as reinforcement for ceramic composites for nuclear applications requires the characterization of its structural stability and mechanical behavior under irradiation. Regarding the radiation stability, ion-amorphization kinetics of these fibers have been studied and compared to the model material, i.e. 6H-SiC single crystals, with no significant differences. For all samples, full amorphization threshold dose yields ∼0.4 dpa at room temperature and complete amorphization was not achieved for irradiation temperatures over 200 C. Successively, ion-amorphized samples have been thermally annealed. It is reported that thermal annealing at high temperatures not only induces the recrystallization of the ion-amorphized samples but also causes unrecoverable mechanical failure, i.e. cracking and delamination. Cracking is reported to be a thermally driven phenomenon characterized by activation energy of 1.05 eV. Regarding the mechanical irradiation behavior, irradiation creep of TSA3 fibers has been investigated using a tensile device dedicated to in situ tests coupled to two different ion-irradiation lines. It is reported that ion irradiation (12 MeV C 4+ and 92 MeV Xe 23+ ) induces a time-dependent strain under loads where thermal creep is negligible. In addition, irradiation strain is reported to be higher at low irradiation temperatures due to a coupling between irradiation swelling and irradiation creep. At high temperatures, near 1000 C, irradiation swelling is minimized hence allowing the characterization of the irradiation creep. Irradiation creep rate is characterized by a linear correlation between the ion flux and the strain rate and a square root dependence with the applied load. Finally, it has been reported that the higher the electronic energy loss contribution to the stopping regime the higher the irradiation creep of the fiber. (author) [fr

  16. Progress in the development of a SiC{sub f}/SiC creep test

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, M.L.; Lewinsohn, C.A.; Jones, R.H.; Youngblood, G.E.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Hecht, S.L.

    1996-10-01

    An effort is now underway to design an experiment that will allow the irradiation creep behavior of SiC{sub f}/SiC composites to be quantified. Numerous difficulties must be overcome to achieve this goal, including determining an appropriate specimen geometry that will fit their radiation volumes available and developing a fabrication procedure for such a specimen. A specimen design has been selected, and development of fabrication methods is proceeding. Thermal and stress analyses are being performed to evaluate the viability of the specimen and to assist with determining the design parameters. A possible alternate type of creep test is also being considered. Progress in each of these areas is described in this report.

  17. Graphite Formation in Cast Iron

    Science.gov (United States)

    Stefanescu, D. M.

    1985-01-01

    In the first phase of the project it was proven that by changing the ratio between the thermal gradient and the growth rate for commercial cast iron samples solidifying in a Bridgman type furnace, it is possible to produce all types of graphite structures, from flake to spheroidal, and all types of matrices, from ferritic to white at a certain given level of cerium. KC-135 flight experiments have shown that in a low-gravity environment, no flotation occurs even in spheroidal graphite cast irons with carbon equivalent as high as 5%, while extensive graphite flotation occurred in both flake and spheroidal graphite cast irons, in high carbon samples solidified in a high gravity environment. This opens the way for production of iron-carbon composite materials, with high carbon content (e.g., 10%) in a low gravity environment. By using KC-135 flights, the influence of some basic elements on the solidification of cast iron will be studied. The mechanism of flake to spheroidal graphite transition will be studied, by using quenching experiments at both low and one gravity for different G/R ratios.

  18. Low Temperature Creep of Hot-Extruded Near-Stoichiometric NiTi Shape Memory Alloy. Part I; Isothermal Creep

    Science.gov (United States)

    Raj, S. V.; Noebe, R. D.

    2013-01-01

    This two-part paper is the first published report on the long term, low temperature creep of hot-extruded near-stoichiometric NiTi. Constant load tensile creep tests were conducted on hot-extruded near-stoichiometric NiTi at 300, 373 and 473 K under initial applied stresses varying between 200 and 350 MPa as long as 15 months. These temperatures corresponded to the martensitic, two-phase and austenitic phase regions, respectively. Normal primary creep lasting several months was observed under all conditions indicating dislocation activity. Although steady-state creep was not observed under these conditions, the estimated creep rates varied between 10(exp -10) and 10(exp -9)/s. The creep behavior of the two phases showed significant differences. The martensitic phase exhibited a large strain on loading followed by a primary creep region accumulating a small amount of strain over a period of several months. The loading strain was attributed to the detwinning of the martensitic phase whereas the subsequent strain accumulation was attributed to dislocation glide-controlled creep. An "incubation period" was observed before the occurrence of detwinning. In contrast, the austenitic phase exhibited a relatively smaller loading strain followed by a primary creep region, where the creep strain continued to increase over several months. It is concluded that the creep of the austenitic phase occurs by a dislocation glide-controlled creep mechanism as well as by the nucleation and growth of deformation twins.

  19. Creep testing and creep loading experiments on friction stir welds in copper at 75 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Henrik C.M.; Seitisleam, Facredin; Sandstroem, Rolf [Corrosion an d Metals Research Institute, Stockholm (Sweden)

    2007-08-15

    Specimens cut from friction stir welds in copper canisters for nuclear waste have been used for creep experiments at 75 deg C. The specimens were taken from a cross-weld position as well as heat affected zone and weld metal. The parent metal specimens exhibited longer creep lives than the weld specimens by a factor of three in time. They in turn were longer than those for the crossweld and HAZ specimens by an order of magnitude. The creep exponent was in the interval 50 to 69 implying that the material was well inside the power-law breakdown regime. The ductility properties expressed as reduction in area were not significantly different and all the rupture specimens demonstrated values exceeding 80%. Experiments were also carried out on the loading procedure of a creep test. Similar parent metal specimens and test conditions were used and the results show that the loading method has a large influence on the strain response of the specimen.

  20. Destruction of nuclear graphite using closed chamber incineration

    International Nuclear Information System (INIS)

    Senor, D.J.; Hollenberg, G.W.; Morgan, W.C.; Marianowski, L.G.

    1994-01-01

    Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO 2 , thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology

  1. Nondestructive evaluation method on mechanical property change of graphite components in the HTGR by ultrasonic wave propagation with grain/pore microstructure

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro

    2003-01-01

    Oxidation damage is one of the crucial factors to degrade mechanical properties of graphite components in the HTGRs. The oxidation increases the porosity of graphite and, hence, results in degradation. In order to evaluate the oxidation damage at neutron irradiated conditions, a new analytical method by ultrasonic wave propagation characteristics was developed. Irradiation effects, a dimensional change and a pinning of dislocations in crystals, on the propagation characteristics in graphite are taken into consideration in the method. It was shown that an equivalent velocity of the wave in graphite is increased by the irradiation, and that a signal height of a propagated waveform is increased by the irradiation, and it decreases with increasing porosity caused by the oxidation. The Young's modulus for an ideal graphite polycrystals without pore was evaluated by considering the wave velocity in them in order to evaluate the change of the apparent modulus at simultaneous irradiated and oxidized conditions as an application of the developed method. It was also shown that the oxidation-induced change of the modulus is appropriately evaluated by the method, suggesting that it is possible to evaluate the change for the irradiated conditions. It can be said from this study that the developed method is promising to evaluate the oxidation damage on graphite components in the HTGRs by nondestructive way. (author)

  2. Application of creep small punch testing in assessment of creep lifetime

    Czech Academy of Sciences Publication Activity Database

    Dobeš, Ferdinand; Milička, Karel

    510-511, Sp. Iss. (2009), s. 440-443 ISSN 0921-5093. [International Conference of Creep and Fracture of Engineering Materials and Structures /11./. Bad Berneck, 04.05.2008-09.05.2008] R&D Projects: GA AV ČR 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : Small punch tests * Creep * Rupture * Chromium steel * Extrapolation methods Subject RIV: JG - Metallurgy Impact factor: 1.901, year: 2009

  3. Analysis of available creep and creep-rupture data for commercially heat-treated alloy 718

    International Nuclear Information System (INIS)

    Booker, M.K.; Booker, B.L.P.

    1980-03-01

    The Ni-Cr-Fe-Nb alloy 718 is a widely used material in elevated- temperature applications. Currently, it is approved by the American Society of Mechanical Engineers ASME Boiler and Pressure Vessel Code only as a bolting material for elevated-temperature nuclear service. This report presents analyses of available creep and creep-rupture data for commercially heat-treated alloy 718 toward the development of allowable stress levels for this material in general elevated-temperature nuclear service. Available data came from 14 heats of bar, plate, and forging material over the temperature range from 538 to 704 degrees C. The longest rupture time encompassed by the data was almost 87,000 h. Generalized regression analyses were performed to yield an analytical expression for rupture life as a function of stress and temperature. Heat-to-heat variations were accounted for by ''lot-centering'' the data. Effects of different solution heat treatment temperatures (T s ) were accounted for by normalizing the creep stresses to the data for T s = 954 degrees C. Thus, the results are strictly applicable only for material with this solution treatment. Time and strain to tertiary creep were predicted as functions of rupture life. Creep strain-time data were represented by normalization to the time and strain to tertiary creep and development of ''master creep curves.'' The results allow estimation of time-dependent allowable stress per American Society of Mechanical Engineers Code Class N-47, and the creep strain-time relationships can be used to develop isochronous stress-strain curves. 29 refs., 44 figs., 14 tabs

  4. Creep rupture behavior of welded Grade 91 steel

    Energy Technology Data Exchange (ETDEWEB)

    Shrestha, Triratna [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Basirat, Mehdi [Department of Mechanical Engineering, University of Idaho, Moscow, ID 83844 (United States); Alsagabi, Sultan; Sittiho, Anumat [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Charit, Indrajit, E-mail: icharit@uidaho.edu [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Potirniche, Gabriel P. [Department of Mechanical Engineering, University of Idaho, Moscow, ID 83844 (United States)

    2016-07-04

    Creep rupture behavior of fusion welded Grade 91 steel was studied in the temperature range of 600 – 700 °C and at stresses of 50–200 MPa. The creep data were analyzed in terms of the Monkman-Grant relation and Larson-Miller parameter. The creep damage tolerance factor was used to identify the origin of creep damage. The creep damage was identified as the void growth in combination with microstructural degradation. The fracture surface morphology of the ruptured specimens was studied by scanning electron microscopy and deformed microstructure examined by transmission electron microscopy, to further elucidate the rupture mechanisms.

  5. Factors influencing the creep strength of hot pressed beryllium

    International Nuclear Information System (INIS)

    Webster, D.; Crooks, D.D.

    1975-01-01

    The parameters controlling the creep strength of hot pressed beryllium block have been determined. Creep strength was improved by a high initial dislocation density, a coarse grain size, and a low impurity content. The impurities most detrimental to creep strength were found to be aluminum, magnesium, and silicon. A uniform distribution of BeO was found to give creep strength which was inferior to a grain boundary distribution. The creep strength of very high purity, hot isostatically pressed beryllium was found to compare favorably with that of other more commonly used high temperature metals

  6. Influence of the phase composition of refractory materials on creep

    Directory of Open Access Journals (Sweden)

    Terzić A.

    2006-01-01

    Full Text Available In this paper, the relationship between the creeping effect and mineralogical characteristics of the applied binding phase for various refractory materials (high-alumina materials, with high or low impurity content, tar bonded either magnesite or dolomite materials and silicate bonded chrom-magnesite materials is presented. The mechanism of creeping is analyzed and the activation energy for creep for each investigated material is obtained and discussed. All investigated materials are creep sensitive under investigated conditions and have similar activation energies for creep except high-alumina refractories with a low impurity content.

  7. Concrete creep at transient temperature: constitutive law and mechanism

    International Nuclear Information System (INIS)

    Chern, J.C.; Bazant, Z.P.; Marchertas, A.H.

    1985-01-01

    A constitutive law which describes the transient thermal creep of concrete is presented. Moisture and temperature are two major parameters in this constitutive law. Aside from load, creep, cracking, and thermal (shrinkage) strains, stress-induced hygrothermal strains are also included in the analysis. The theory agrees with most types of test data which include basic creep, thermal expansion, shrinkage, swelling, creep at cyclic heating or drying, and creep at heating under compression or bending. Examples are given to demonstrate agreement between the theory and the experimental data. 15 refs., 6 figs

  8. Compressive creep of hot pressed silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Silva, C.R.M., E-mail: cosmeroberto@gmail.com [Universidade de Brasilia (UnB), Campus Darcy Ribeiro, Brasilia CEP 70736-020, DF (Brazil); Nono, M.C.A. [Instituto de Nacional de Pesquisas Espaciais (INPE-LAS) (Brazil); Reis, D.A.P.; Hwang, M.K. [Instituto de Aeronautica e Espaco (IAE) (Brazil)

    2010-07-15

    Silicon carbide has a good match of chemical, mechanical and thermal properties and therefore is considered an excellent structural ceramic for high temperature applications. The aim of the present work is compressive creep evaluation of liquid phase sintered silicon carbide with aluminum and rare earth oxide as sintering aids. Rare earth oxides are possible additives considering their highly refractory remnant grain-boundary phase and lower synthesis costs compared to high purity rare earth. Samples were prepared with silicon carbide powder (90 wt%) and aluminum oxide (5 wt%) plus rare earth oxide (5 wt%) additions. Powders were mixed, milled and hot pressed at 1800 deg. C in argon atmosphere. Compressive creep tests were carried out under stress from 150 to 300 MPa and temperatures from 1300 to 1400 deg. C. At lower creep test temperatures, the obtained stress exponent values were correlated to mechanisms based on diffusion. At intermediate temperatures, grain-boundary sliding becomes operative, accommodated by diffusion. At higher temperatures cavities are discernible. Oxidation reactions and ionic diffusion result on surface oxidized layer, grain-boundary amorphous and intergranular crystalline Al{sub 6}Si{sub 2}O{sub 13}, {delta}-Y{sub 2}Si{sub 2}O{sub 7} and YAG phases. In this case cavitation and amorphous phases redistribution enhance grain-boundary sliding, not accommodated by diffusion. Coalescence occurs at triple point and multigrain-junctions, with subsequent strain rate acceleration and cavitational creep.

  9. Viscous creep in metals at intermediate temperatures

    Czech Academy of Sciences Publication Activity Database

    Kloc, Luboš; Fiala, J.

    2005-01-01

    Roč. 43, č. 2 (2005), s. 105-112 ISSN 0023-432X R&D Projects: GA AV ČR(CZ) IAA2041101 Institutional research plan: CEZ:AV0Z20410507; CEZ:AV0Z2041904 Keywords : creep * heat resistant steel Subject RIV: JG - Metallurgy Impact factor: 0.973, year: 2005

  10. Creep measurements on curing epoxy systems

    DEFF Research Database (Denmark)

    Kammer, Charlotte; Szabo, Peter

    1998-01-01

    The chemical curing of a stoichiometric mixture of the diglycidyl ether of bisphenol A and a 1,3-bis-(aminomethyl)-cyclohexane is studied.Creep experiments are combined with measurements in a Differential Scanning Calorimeter (DSC) to determine the change in bulk viscosity due to network formation....

  11. Creep of granulated loose-fill insulation

    DEFF Research Database (Denmark)

    Rasmussen, Torben Valdbjørn

    with SP-Building Physics in Sweden and VTT Building Technology in Finland. For the round robin test a cellulosic fibre insulation material was used. The proposed standardised method for creep tests and theories are limited to cases when the granulated loose-fill material is exposed to a constant......, Organisation for Testing in Scandinavia funded the Nordtest....

  12. Plasticity - a limiting case of creep

    International Nuclear Information System (INIS)

    Cords, H.; Kleist, G.; Zimmermann, R.

    1986-11-01

    The present work is an attempt to develop further the so-called unified theory for viscoplastic constitutive equations as used for metals or metal alloys. Typically, in similar approaches creep strains and plastic strains are derived from one common stress-strain relationship for inelastic strain rates employing an internal stress function as a back stress. Some novel concepts concerning the definition of the internal stress, plastic yielding and material hardening have been introduced, formulated mathematically and tested for correspondence with a standard type of materials behaviour. As a result of the investigations a system of simultaneous differential equations is defined which has been used to elaborate a common view on a number of different material effects observed in creep and plasticity i.e. normal and inverted primary creep, recoverable creep, incubation time and anelasticity in stress reduction, negative stress relaxation, plastic yielding, perfect plasticity, negative strain rate sensitivity, serrated flow, strain hardening in monotonic and cyclic loading. The theoretical approach is mainly based on a lateral contraction movement not following rigidly the longitudinal extension of the material specimen by a prescribed constant value of Poisson's ratio as usual, but following the axial extension in a process of drag which allows for retardation and which simultaneously impedes the longitudinal straining. (orig.) [de

  13. Creep properties of aluminium processed by ECAP

    Czech Academy of Sciences Publication Activity Database

    Král, Petr; Dvořák, Jiří; Jäger, Aleš; Kvapilová, Marie; Horita, Z.; Sklenička, Václav

    2016-01-01

    Roč. 54, č. 6 (2016), s. 441-451 ISSN 0023-432X R&D Projects: GA MŠk(CZ) LQ1601 Institutional support: RVO:68081723 ; RVO:68378271 Keywords : equal channel angular pressing (ECAP) * aluminium * ultrafine-grained microstructure * creep Subject RIV: JG - Metallurgy; JG - Metallurgy (FZU-D) Impact factor: 0.366, year: 2016

  14. Constant structure creep experiments on aluminium

    Czech Academy of Sciences Publication Activity Database

    Milička, Karel

    2011-01-01

    Roč. 49, č. 5 (2011), s. 307-318 ISSN 0023-432X R&D Projects: GA AV ČR IAA2041203 Institutional research plan: CEZ:AV0Z20410507 Keywords : mechanical properties * high temperature deformation * creep * aluminium Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.451, year: 2011

  15. timber joists subjected to creep-rupture

    African Journals Online (AJOL)

    user

    Wood experiences a significant loss of strength and stiffness when loaded over period of time. This phenomenon is known as creep-rupture. Several models were developed for the estimation of the reduction of load carrying capacity of timber with time. In this paper, the results of time dependent structural reliability analysis ...

  16. Energy response of graphite-mixed magnesium borate TLDs to low energy x-rays

    DEFF Research Database (Denmark)

    Pelliccioni, M.; Prokic, M.; Esposito, A.

    1991-01-01

    Graphite-mixed sintered magnesium borate TL dosemeters are attractive for beta/gamma dosimetry because they combine a low energy dependence to beta-rays with near tissue or air equivalence to photon irradiations and a high sensitivity. In this paper results from the experimental measurements...

  17. A constitutive equation of creep based on recoverable creep hardening range

    International Nuclear Information System (INIS)

    Murakami, Sumio; Ohno, Nobutada

    1982-01-01

    In case of the stress reverse test of 304 stainless steel under constant stress, the creep curve after stress reverse is not much different from that for the initial state, accordingly, the state of hardening of the material is restored almost to that in the initial state by stress reverse. Creep depends on past loading history, and essentially is anisotropic phenomenon. In this study, by extending the strain hardening theory to general multi-axial stress condition, a constitutive formula for creep without any difficulty was determined. For the purpose, as the main change of microscopic structure at the time of stress reverse in mono-axial creep, the effect of the reactivation of passive dislocations was considered. At this time, the range of recoverable creep hardening was assumed, in which irreversible dislocation arrangement does not occur after stress reverse. In order to examine the propriety of this theory, the multi-axial creep deformation, of which the principal stress direction changes periodically, was calculated by the derived constitutive formula, and the result was compared with the experimental result of 304 stainless steel tube at 650 deg C and the theoretical result of ORNL. (Kako, I.)

  18. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  19. Mechanical properties of graphites and carbon materials

    International Nuclear Information System (INIS)

    Jouquet, Gilbert.

    1977-01-01

    The mechanical behavior of graphites and artificial carbons is related to the structure of these materials. The influence of structural modifications in a graphite monocrystal on the deformation and fracture properties is studied [fr

  20. Uniaxial creep behavior of V-4Cr-4Ti alloy

    International Nuclear Information System (INIS)

    Natesan, K.; Soppet, W.K.; Purohit, A.

    2002-01-01

    We are undertaking a systematic study at Argonne National Laboratory to evaluate the uniaxial creep behavior of V-Cr-Ti alloys in a vacuum environment as a function of temperature in the range of 650-800 deg. C and at applied stress levels of 75-380 MPa. Creep strain in the specimens is measured by a linear-variable-differential transducer, which is attached between the fixed and movable pull rods of the creep assembly. Strain is measured at sufficiently frequent intervals during testing to define the creep strain/time curve. A linear least-squares analysis function is used to ensure consistent extraction of minimum creep rate, onset of tertiary creep and creep strain at the onset of tertiary creep. Creep test data, obtained at 650, 700, 725 and 800 deg. C, showed power-law creep behavior. Extensive analysis of the tested specimens is conducted to establish hardness profiles, oxygen content and microstructural characteristics. The data are also quantified by the Larson-Miller approach, and correlations are developed to relate time to rupture, onset of tertiary creep, times for 1% and 2% strain, exposure temperature and applied stress

  1. Nonlinear creep damage constitutive model for soft rocks

    Science.gov (United States)

    Liu, H. Z.; Xie, H. Q.; He, J. D.; Xiao, M. L.; Zhuo, L.

    2017-02-01

    In some existing nonlinear creep damage models, it may be less rigorous to directly introduce a damage variable into the creep equation when the damage variable of the viscous component is a function of time or strain. In this paper, we adopt the Kachanov creep damage rate and introduce a damage variable into a rheological differential constitutive equation to derive an analytical integral solution for the creep damage equation of the Bingham model. We also propose a new nonlinear viscous component which reflects nonlinear properties related to the axial stress of soft rock in the steady-state creep stage. Furthermore, we build an improved Nishihara model by using this new component in series with the correctional Nishihara damage model that describes the accelerating creep, and deduce the rheological constitutive relation of the improved model. Based on superposition principle, we obtain the damage creep equation for conditions of both uniaxial and triaxial compression stress, and study the method for determining the model parameters. Finally, this paper presents the laboratory test results performed on mica-quartz schist in parallel with, or vertical to the schistosity direction, and applies the improved Nishihara model to the parameter identification of mica-quartz schist. Using a comparative analysis with test data, results show that the improved model has a superior ability to reflect the creep properties of soft rock in the decelerating creep stage, the steady-state creep stage, and particularly within the accelerating creep stage, in comparison with the traditional Nishihara model.

  2. Creep-fatigue damage assessment by subsequent fatigue straining

    International Nuclear Information System (INIS)

    Yaguchi, M.; Nakamura, T.; Ishikawa, A.; Asada, Y.

    1993-01-01

    A series of creep-fatigue tests has been conducted with Modified 9Cr-1Mo steel at 600 deg. C in a high vacuum environment of 0.1mPa to assess an accumulation of creep-fatigue damage. In these tests, each test specimen has been subjected to prior creep-fatigue loading followed by subsequent fatigue loading or prior fatigue loading followed by subsequent creep-fatigue loading. A linear summation of cumulative damage of fatigue and creep life fraction is smaller than unity for the former case, and larger than unity for the latter case. SEM observation was conducted and it was shown that in the case of prior creep-fatigue loading, crack mode transforms from transgranular to intergranular type with the increase of the number of cycles of prior creep-fatigue loading, while crack mode is generally intergranular in the case of prior fatigue loading. (author)

  3. Soil creep and historic landscape changes

    Science.gov (United States)

    Lucke, Bernhard

    2017-04-01

    Many erosion models assume that soil sediments are transported grain-by-grain, and thus calculate loss and deposition according to parameters such as bulk density and average grain size. However, clay-rich soils, such as the widespread Red Mediterranean Soils or Terrae Rossae that are often found near important archaeological sites, can behave differently. This is illustrated by a case study of historic landscape changes in Jordan, where evidence for soil creep as main process of soil movement was found in the context of ancient cemeteries. Due to a dominance of smectites, the Red Mediterranean Soils in this area shrink and form cracks during the dry period. Because of the cracks and underlying limestone karst, they can swallow strong rains without high erosion risk. However, when water-saturated, these soils expand and can start creeping. Buried geoarchaeological features like small water channels on formerly cleared rocks suggest that soils can move a few cm uplslope when wet, and buried graves illustrate that soil creep can create new level surfaces, sealing cavities but not completely filling them. Such processes seem associated with slumping and earth flows as instable rocks might collapse under the weight of a creeping soil. While it is very difficult to measure such processes, landscape archaeology offers at least an indirect approach that could be suited to estimate the scale and impact of soil creep. Analogies with modern rainfalls, including record levels of precipitation during the winter 1991/1992, indicate that similar levels of soil moisture have not been reached during times of modern instrumental rainfall monitoring. This suggests that very strong deluges must have occurred during historical periods, that could potentially cause tremendous damage to modern infrastructure if happening again.

  4. RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

    Science.gov (United States)

    Fromm, L.W. Jr.

    1959-09-01

    An electrolytic process for recovering uranium from a graphite fuel element is described. The uraniumcontaining graphite body is disposed as the anode of a cell containing a nitric acid electrolyte and a 5 amp/cm/sup 2/ current passed to induce a progressive disintegration of the graphite body. The dissolved uranium is quickly and easily separated from the resulting graphite particles by simple mechanical means, such as centrifugation, filtration, and decontamination.

  5. Residual stress measurements in polycrystalline graphite with micro-Raman spectroscopy

    International Nuclear Information System (INIS)

    Krishna, Ram; Jones, Abbie N.; Edge, Ruth; Marsden, Barry J.

    2015-01-01

    Micro-Raman microscopy technique is applied to evaluate unevenly distributed residual stresses in the various constituents of polygranular reactor grades graphite. The wavenumber based Raman shift (cm −1 ) corresponds to the local residual stress and measurements of stress dependent first order Raman spectra in graphite have enabled localized residual stress values to be determined. The bulk polygranular graphite of reactor grades – Gilsocarbon, NBG-18 and PGA – are examined to illustrate the residual stress variations in their constituents. Binder phase and filler particles have shown to be under compressive and tensile stresses, respectively. Among the studied graphite grades, the binder phase in Gilsocarbon has the highest residual stress and NBG-18 has the lowest value. Filler particles in Gilsocarbon have the highest residual stress and PGA showed the lowest, this is most likely due to the morphology of the coke particles used in the manufacturing and applied processing techniques for fabrications. Stresses have also been evaluated along the peripheral of pores and at the tips of the cracks. Cracks in filler and binder phases have shown mixed behaviour, compressive as well as tensile, whereas pores in binder and filler particles have shown compressive behaviour. The stresses in these graphitic constituents are of the order of MPa. Non-destructive analyses presented in this study make the current state-of-the-art technique a powerful method for the study of stress variations near the graphite surface and are expected to increase its use further in property determination analysis of low to highly fluence irradiated graphite samples from the material test reactors. - Highlights: • Micro-Raman spectroscopy can measure significantly small residual stresses. • Gilsocarbon, NBG-18 and PGA graphite were evaluated for residual stresses. • Residual stresses in the constituents of graphite were evaluated. • Binder and filler particles are often found under

  6. Characterization of radiation damage induced by swift heavy ions in graphite

    Energy Technology Data Exchange (ETDEWEB)

    Hubert, Christian

    2016-05-15

    Graphite is a classical material in neutron radiation environments, being widely used in nuclear reactors and power plants as a moderator. For high energy particle accelerators, graphite provides ideal material properties because of the low Z of carbon and its corresponding low stopping power, thus when ion projectiles interact with graphite is the energy deposition rather low. This work aims to improve the understanding of how the irradiation with swift heavy ions (SHI) of kinetic energies in the range of MeV to GeV affects the structure of graphite and other carbon-based materials. Special focus of this project is given to beam induced changes of thermo-mechanical properties. For this purpose the Highly oriented pyrolytic graphite (HOPG) and glassy carbon (GC) (both serving as model materials), isotropic high density polycrystalline graphite (PG) and other carbon based materials like carbon fiber carbon composites (CFC), chemically expanded graphite (FG) and molybdenum carbide enhanced graphite composites (MoC) were exposed to different ions ranging from {sup 131}Xe to {sup 238}U provided by the UNILAC accelerator at GSI in Darmstadt, Germany. To investigate structural changes, various in-situ and off-line measurements were performed including Raman spectroscopy, x-ray diffraction and x-ray photo-electron spectroscopy. Thermo-mechanical properties were investigated using the laser-flash-analysis method, differential scanning calorimetry, micro/nano-indentation and 4-point electrical resistivity measurements. Beam induced stresses were investigated using profilometry. Obtained results provided clear evidence that ion beam-induced radiation damage leads to structural changes and degradation of thermal, mechanical and electrical properties of graphite. PG transforms towards a disordered sp2 structure, comparable to GC at high fluences. Irradiation-induced embrittlement is strongly reducing the lifetime of most high-dose exposed accelerator components. For

  7. A physics-based crystallographic modeling framework for describing the thermal creep behavior of Fe-Cr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wen, Wei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, Laurent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    This Report addresses the Milestone M2MS-16LA0501032 of NEAMS Program (“Develop hardening model for FeCrAl cladding), with a deadline of 09/30/2016. Here we report a constitutive law for thermal creep of FeCrAl. This Report adds to and complements the one for Milestone M3MS-16LA0501034 (“Interface hardening models with MOOSE-BISON”), where we presented a hardening law for irradiated FeCrAl. The last component of our polycrystal-based constitutive behavior, namely, an irradiation creep model for FeCrAl, will be developed as part of the FY17 Milestones, and the three regimes will be coupled and interfaced with MOOSE-BISON.

  8. Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

    International Nuclear Information System (INIS)

    Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

    2014-01-01

    Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

  9. Finite element creep buckling analysis of circular cylindrical shell under axial compression taking account of creep damage

    Science.gov (United States)

    Hagihara, Seiya; Miyazaki, Noriyuki

    1998-05-01

    Cylindrical shells are utilized as structural elements of nuclear power plants, heat exchangers or pressure vessels, which are operated under elevated temperature. Creep buckling is one of the failure modes of structures at elevated temperature. In some experiments conducted by other authors, axially compressive cylindrical shells with a large ratio of radius to thickness were observed to buckle with circumferential waves. It is observed that the circumferential waves occur due to bifurcation buckling. But, the critical time and the minimum loading for bifurcation buckling obtained from calculations of finite element analyses are not in very good agreement with those of the experiments. One of the reasons for the disagreement is considered to be that the creep constitutive equations employed in many previous analyses represent the steady creep. The creep phenomena usually have primary creep period, steady creep one and tertiary creep one. A creep strain - time relation through the three periods can be simulated by using a constitutive equation based on creep damage mechanics. In the present analysis, we analyzed the bifurcation creep buckling of circular cylindrical shells subjected to axial compression by the use of the finite element method taking account of the creep damage mechanics proposeol by of Kachanov-Rabotonov.

  10. Superconductivity in graphite intercalation compounds

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Robert P. [Cavendish Laboratory, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Weller, Thomas E.; Howard, Christopher A. [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom); Dean, Mark P.M. [Department of Condensed Matter Physics and Materials Science, Brookhaven National Laboratory, Upton, NY 11973 (United States); Rahnejat, Kaveh C. [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom); Saxena, Siddharth S. [Cavendish Laboratory, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Ellerby, Mark, E-mail: mark.ellerby@ucl.ac.uk [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom)

    2015-07-15

    Highlights: • Historical background of graphite intercalates. • Superconductivity in graphite intercalates and its place in the field of superconductivity. • Recent developments. • Relevant modeling of superconductivity in graphite intercalates. • Interpretations that pertain and questions that remain. - Abstract: The field of superconductivity in the class of materials known as graphite intercalation compounds has a history dating back to the 1960s (Dresselhaus and Dresselhaus, 1981; Enoki et al., 2003). This paper recontextualizes the field in light of the discovery of superconductivity in CaC{sub 6} and YbC{sub 6} in 2005. In what follows, we outline the crystal structure and electronic structure of these and related compounds. We go on to experiments addressing the superconducting energy gap, lattice dynamics, pressure dependence, and how these relate to theoretical studies. The bulk of the evidence strongly supports a BCS superconducting state. However, important questions remain regarding which electronic states and phonon modes are most important for superconductivity, and whether current theoretical techniques can fully describe the dependence of the superconducting transition temperature on pressure and chemical composition.

  11. Graphite nanoreinforcements in polymer nanocomposites

    Science.gov (United States)

    Fukushima, Hiroyuki

    Nanocomposites composed of polymer matrices with clay reinforcements of less than 100 nm in size, are being considered for applications such as interior and exterior accessories for automobiles, structural components for portable electronic devices, and films for food packaging. While most nanocomposite research has focused on exfoliated clay platelets, the same nanoreinforcement concept can be applied to another layered material, graphite, to produce nanoplatelets and nanocomposites. Graphite is the stiffest material found in nature (Young's Modulus = 1060 GPa), having a modulus several times that of clay, but also with excellent electrical and thermal conductivity. The key to utilizing graphite as a platelet nanoreinforcement is in the ability to exfoliate this material. Also, if the appropriate surface treatment can be found for graphite, its exfoliation and dispersion in a polymer matrix will result in a composite with not only excellent mechanical properties but electrical properties as well, opening up many new structural applications as well as non-structural ones where electromagnetic shielding and high thermal conductivity are requirements. In this research, a new process to fabricate exfoliated nano-scale graphite platelets was established (Patent pending). The size of the resulted graphite platelets was less than 1 um in diameter and 10 nm in thickness, and the surface area of the material was around 100 m2/g. The reduction of size showed positive effect on mechanical properties of composites because of the increased edge area and more functional groups attached with it. Also various surface treatment techniques were applied to the graphite nanoplatelets to improve the surface condition. As a result, acrylamide grafting treatment was found to enhance the dispersion and adhesion of graphite flakes in epoxy matrices. The resulted composites showed better mechanical properties than those with commercially available carbon fibers, vapor grown carbon fibers

  12. Study of high dose nitrogen implantation into graphite

    International Nuclear Information System (INIS)

    Romanovskij, E.A.; Bespalova, O.V.; Borisov, A.M.; Goryaga, N.G.; Zatekin, V.V.; Kulikauskas, V.S.; Sukharev, V.G.

    1997-01-01

    Rutherford backscattering spectroscopy was used for the study of high dose (35 keV)N + ions implantation into graphites and glassy carbon. Quantitative date on depth profiles and its dependences on irradiation fluence and ion beam flux were obtained for all elements. The stationary cupola-shaped depth profile with maximum nitrogen concentration 22-27% (at.) is reached at sufficiently large fluence. The obtained results are discussed in the frame of high dose implantation models and compared with results of another methods of carbon nitride synthesis

  13. Influence of phosphorus on the creep ductility of copper

    International Nuclear Information System (INIS)

    Sandström, Rolf; Wu, Rui

    2013-01-01

    Around 1990 it was discovered that pure copper could have extra low creep ductility in the temperature interval 180–250 °C. The material was intended for use in canisters for nuclear waste disposal. Although extra low creep ductility was not observed much below 180 °C and the temperature in the canister will never exceed 100 °C, it was feared that the creep ductility could reach low values at lower temperatures after long term exposure. If 50 ppm phosphorus was added to the copper the low creep ductility disappeared. A creep cavitation model is presented that can quantitatively describe the cavitation behaviour in uniaxial and multiaxial creep tests as well as the observed creep ductility for copper with and without phosphorus. A so-called double ledge model has been introduced that demonstrates why the nucleation rate of creep cavities is often proportional to the creep rate. The phosphorus agglomerates at the grain boundaries and limits their local deformation and thereby reduces the formation and growth of cavities. This explains why extra low creep ductility does not occur in phosphorus alloyed copper

  14. An anisotropic tertiary creep damage constitutive model for anisotropic materials

    International Nuclear Information System (INIS)

    Stewart, Calvin M.; Gordon, Ali P.; Ma, Young Wha; Neu, Richard W.

    2011-01-01

    When an anisotropic material is subject to creep conditions and a complex state of stress, an anisotropic creep damage behavior is observed. Previous research has focused on the anisotropic creep damage behavior of isotropic materials but few constitutive models have been developed for anisotropic creeping solids. This paper describes the development of a new anisotropic tertiary creep damage constitutive model for anisotropic materials. An advanced tensorial damage formulation is implemented which includes both material orientation relative to loading and the degree of creep damage anisotropy in the model. A variation of the Norton-power law for secondary creep is implemented which includes the Hill's anisotropic analogy. Experiments are conducted on the directionally-solidified bucket material DS GTD-111. The constitutive model is implemented in a user programmable feature (UPF) in ANSYS FEA software. The ability of the constitutive model to regress to the Kachanov-Rabotnov isotropic tertiary creep damage model is demonstrated through comparison with uniaxial experiments. A parametric study of both material orientation and stress rotation are conducted. Results indicate that creep deformation is modeled accurately; however an improved damage evolution law may be necessary. - Highlights: → The deformation of anisotropic creeping solid is directionally dependent. → Few constitutive models have been developed to deal with anisotropic behavior. → A transversely-isotropic nickel base superalloy, DS GTD-111, is studied. → A vector constitutive model based on the Kachanov-Rabotnov formulation is developed. → The new model accurately models deformation at various orientations.

  15. Laser ultrasonic assessment of the effects of porosity and microcracking on the elastic moduli of nuclear graphites

    International Nuclear Information System (INIS)

    Spicer, James B.; Olasov, Lauren R.; Zeng, Fan W.; Han, Karen; Gallego, Nidia C.; Contescu, Cristian I.

    2016-01-01

    Laser ultrasonic methods have been used to measure the elastic moduli of various nuclear graphites. Measurements were made to assess wavespeeds for longitudinal and shear waves in different propagation directions and these were used along with density measurements to compute the longitudinal and shear moduli as well as Young's modulus. All moduli decreased with increasing graphite porosity and these variations could be interpreted using models describing the effect of porosity on material stiffness. Extrapolations for these models to zero porosity were used to infer the moduli for theoretically dense graphite; the results were far below predicted values reported in the literature for fully dense, polycrystalline, isotropic graphite. Differences can be attributed to microcracking in the graphite microstructure. Using models for the effects of microcracking on modulus, estimates for microcrack populations indicate that the number of cracks per unit volume must be much greater than the number of pores per unit volume. Experimental results reported in the literature for irradiated graphites as well as for the stress dependence of graphite modulus are consistent with the influence of microcracking on elastic behavior and could be interpreted using concepts developed here. Results in this work for graphite structure-property relationships should allow for more sophisticated characterization of nuclear graphites using ultrasonic methods. - Highlights: • Moduli of nuclear graphites measured using laser ultrasonic methods. • Estimates made for the moduli of fully dense, polycrystalline,